Symposium Organizers
Katherine L. Smith Embassy of Australia
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Scott Kroeker University of Manitoba
Blas Uberuaga Los Alamos National Laboratory
AA1: Repositories and National Programs
Session Chairs
Tuesday PM, April 06, 2010
Room 3010 (Moscone West)
9:30 AM - **AA1.1
Applying Insights from Repository Safety Assessments.
Peter Swift 1
1 Organization 06780, Sandia National Laboratories, Albuquerque, New Mexico, United States
Show AbstractDespite decades of international consensus that deep geological disposal is the best option for permanent management of long-lived high-level radioactive wastes, no repositories for used nuclear fuel or high-level waste are in operation. Detailed long-term safety assessments have been completed worldwide for a wide range of repository designs and disposal concepts, however, and valuable insights from these assessments are available to inform future decisions about managing radioactive wastes. Qualitative comparisons among the existing safety assessments for disposal concepts in clay, granite, salt, and unsaturated volcanic tuff show how different geologic settings can be matched with appropriate engineered barrier systems to provide a high degree of confidence in the long-term safety of geologic disposal. Review of individual assessments provides insights regarding the release pathways and radionuclides that are most likely to contribute to estimated doses to humans in the far future for different disposal concepts, and can help focus research and development programs to improve management and disposal technologies. Lessons learned from existing safety assessments may be particularly relevant for informing decisions during the process of selecting potential repository sites. This abstract is Sandia publication SAND2009-8065A. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
10:00 AM - **AA1.2
Fuel Cycle Research and Development Program, Used Fuel Disposition Campaign Objective, Mission, Plans, and Activity Status.
W. Mark Nutt 1
1 , Argonne National Laboratory, Argonne, Illinois, United States
Show AbstractThe safe management and disposition of used nuclear fuel and/or high level nuclear waste is a fundamental aspect of the nuclear fuel cycle. The United States currently utilizes a once-through fuel cycle where used nuclear fuel is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. However, a decision not to use the proposed Yucca Mountain Repository will result in longer interim storage at reactor sites than previously planned. In addition, alternatives to the once-through fuel cycle are being considered and a variety of options are being explored under the U.S. Department of Energy’s Fuel Cycle Research and Development Program. These two factors lead to the need to develop a credible strategy for managing radioactive wastes from any future nuclear fuel cycle in order to provide acceptable disposition pathways for all wastes regardless of transmutation system technology, fuel reprocessing scheme(s), and/or the selected fuel cycle. These disposition paths will involve both the storing of radioactive material for some period of time and the ultimate disposal of radioactive waste. As disposition paths evolve from the continuing research and development process, it is important that storage options for fuel cycle materials remain as flexible as possible in order to facilitate selected disposal options.The disposal of radioactive waste of all classifications (low-, intermediate-, high-level waste, and used nuclear fuel) has been investigated world-wide since the inception of nuclear power. While significant progress has been made regarding disposal, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Experience with the Yucca Mountain Project has illustrated the challenges of siting, characterizing, designing, and licensing of a geologic repository. Progress has been demonstrated by the deployment of near-surface disposal facilities for low level waste and the Waste Isolation Pilot Plant for the disposal of defense-related transuranic wastes.However, the capacity for disposing low level wastes is limited, potential disposal pathways for Greater Than Class C low level waste (which is essentially intermediate level waste) have yet to be identified, and the disposal of used nuclear fuel has not been demonstrated. An expansion of nuclear power in the United States, and world-wide could increase the amount of all classes of waste and requires the availability of routine disposal pathways, and adoption of new fuel cycle strategies could result in new requirements for storage and disposal.To address these challenges, the DOE Office of Nuclear Energy established the Used Fuel Disposition Campaign within its Fuel Cycle Research and Development Program in the summer of 2009. The mission of the Used Fuel Disposition Campaign is to identify alternatives and conduct scientific research and technology development to enable storage and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The near-and long-term plans of the Used Fuel Disposition Campaign will be presented and the results of on-going activities being conducted under the campaign will be summarized.
10:30 AM - AA1.3
Establishment of Uncertainty Ranges and Probability Distributions of Actinide Solubilities for Performance Assessment in the Waste Isolation Pilot Plant.
Yongliang Xiong 1
1 , Sandia National Laboratories, Carlsbad, New Mexico, United States
Show AbstractThe Fracture-Matrix Transport (FMT) code developed at Sandia National Laboratories (Novak, 1996; Babb and Novak, 1997 and addenda; Wang, 1998; Giambalvo et al., 2002; Xiong et al., 2005) solves chemical equilibrium problems using the Pitzer activity coefficient model with a database containing actinide species. The code is capable of predicting actinide solubilities at 25 oC in various ionic-strengh solutions from dilute groundwaters to high-ionic-strength brines. The code uses oxidation state analogies, i.e., Am(III) is used to predict solubilities of actinides in the +III oxidation state; Th(IV) is used to predict solubilities of actinides in the +IV state; Np(V) is utilized to predict solubilities of actinides in the +V state. This code has been qualified for predicting actinide solubilities for the Waste Isolation Pilot Plant (WIPP) Compliance Certification Application in 1996, and Compliance Re-Certification Applications in 2004 and 2009. We have established revised actinide-solubility uncertanity ranges and probability distributions for Performance Assessment (PA) by comparing actinide solubilities predicted by FMT with solubility data in various solutions from the open literature. The literature data used in this study includes solubilities in simple solutions (NaCl, NaHCO3, Na2CO3, NaClO4, KCl, K2CO3, etc.), binary solutions (NaCl+NaHCO3, NaCl+Na2CO3, KCl+K2CO3, etc.), ternary solutions (NaCl+Na2CO3+KCl, NaHCO3+Na2CO3+NaClO4, etc.), and multi-component synthetic brines relevant to the WIPP. This research is funded by WIPP programs administered by the U.S. Department of Energy. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. References[1] Babb, S.C., and C.F. Novak. (1997), “User’s Manual for FMT Version 2.3: A Computer Code Employing the Pitzer Activity Coefficient Formalism for Calculating Thermodynamic Equilibrium in Geochemical Systems to High Electrolyte Concentrations.” Albuquerque, NM: Sandia National Laboratories. [2] Giambalvo E., Brush L.H., Xiong Y.-L, (2002), Eos Trans. AGU, 83 (47), Fall Meet. Suppl., Abstract U11B-02, 2002.[3] Novak C.F., (1996), J. Contaminant Hydrology, 21, 297-310.[4] Wang, Y.-F., (1998), “WIPP PA Validation Document for FMT (Version 2.4), Document Version 2.4.” Carlsbad, NM: Sandia National Laboratories. [5] Xiong, Y.-L., Nowak, E.J. and Brush, L.H., (2005), Geochimica et Cosmochimica Acta, 69(10), Supp.1, A417.
10:45 AM - AA1.4
Uranium Chemistry in the Waste Isolation Pilot Plant.
Jean-Francois Lucchini 1 , Hnin Khaing 1 , Marian Borkowski 1 , Michael Richmann 1 , Juliet Swanson 1 , Donald Reed 1
1 EES-12, Los Alamos National Laboratory, Carlsbad, New Mexico, United States
Show AbstractWhen present, uranium is usually an element of importance in a nuclear waste repository. In the Waste Isolation Pilot Plant (WIPP), uranium is the most prevalent actinide component by mass, with about 647 metric tons to be placed in the repository [1]. Therefore, the chemistry of uranium, and especially its solubility in the WIPP conditions, needs to be well determined. In the anoxic and strongly reducing environment expected in the WIPP, tetravalent uranium will be the dominant oxidation state. As a consequence, the uranium solubility will be very low (about 10-8M). However, some uranium (VI) phases and aqueous species, although not expected to predominate in the WIPP, could be present due to the localized effects of radiolysis. The presence of hexavalent uranium will potentially increase the overall uranium solubility in the repository. Long-term experiments were performed to establish the uranium (VI) solubility limits in WIPP brine, and to evaluate the contribution of carbonate complexation and hydrolysis to uranium (VI) speciation. Even in the presence of carbonate (at millimole levels), experimental results showed that uranium (VI) concentrations will not exceed 10-4M. This measured solubility limit is an order of magnitude lower than the uranium solubility value currently used in the WIPP Performance Assessment (PA) [1]. The WIPP PA also considers that uranium will speciate as U(VI) with a probability of 0.5 in the PA vectors [1]. This is a conservative assumption, considering the reducing conditions expected in the WIPP. Sustainability of uranium (VI) in the WIPP was then challenged with laboratory experiments, to explore possible reduction pathways in the WIPP such as iron reduction and bioreduction.This paper will address the major expected aspects of the uranium chemistry in the WIPP, and summarize our experimental results to establish its likely speciation under WIPP-relevant conditions. It will include uranium speciation, solubility, complexation and reduction.[1] : U.S. Department of Energy (DOE). 2009. Title 40 CFR Part 191 Subparts B and C Compliance Recertification Application for the Waste Isolation Pilot Plant. Appendix SOTERM-2009. DOE/WIPP 2009-3424. Carlsbad, NM: Carlsbad Field Office.
11:30 AM - **AA1.5
Immobilization of Fission Products in Complex Oxides: Example: (Ln)2Tc2O7 Pyrochlore to Immobilize Tc-99.
Kurt Sickafus 1 , Thomas Hartmann 2 , Phil Weck 2 , Chao Jiang 1 , Frederic Poineau 2 , Ken Czerwinski 2 , Gordon Jarvinen 1 , James Valdez 1 , Blas Uberuaga 1
1 Materials Sciences Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Department of Chemistry, Unversity of Nevada, Las Vegas, Las Vegas, Nevada, United States
Show AbstractWe have synthesized several pyrochlore-structured complex oxides, intended as host materials for the sequestration of the long-lived radiotoxic fission product, technetium (Tc) 99. Specifically, we synthesized (Ln)2Tc2O7 compounds using five different lanthanides (Ln): Pr, Nd, Sm, Gd, and Lu. We performed X-ray diffraction and crystal structure Rietveld refinements, in order to quantify the cubic lattice parameter, a, for each compound, as well as the degree of cation order and the oxygen parameter, x. We also performed density functional theory (DFT) calculations in which we determined theoretical values for a and x in fully-ordered (Ln)2Tc2O7 compounds. In this presentation, we will compare and contrast the experimental and theoretical results described above. We will particularly examine changes in ionic partial charge as the Ln species is varied in our (Ln)2Tc2O7 pyrochlore compounds.
12:00 PM - **AA1.6
Nuclear Waste Management in the United States – Lessons Learned.
Rodney Ewing 1
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractOn January 7, 1983, President Reagan signed the Nuclear Waste Policy Act of 1982. During the following years, the NWPA was modified or amended in 1987 and 1992. In 2002, President Bush recommended the Yucca Mountain site in Nevada as a geologic repository for spent nuclear fuel (SNF) and high-level waste (HLW). The license application was submitted to the Nuclear Regulatory Commission (NRC) on June 3, 2008. The development of a standard by the Environmental Protection Agency and the implementing regulations by the NRC required a parallel, 25-year effort, punctuated by recommendations on the standard from the National Research Council in 1995 and periodic rulings in federal courts. In 2009, the Obama administration announced that Yucca Mountain would not become a geologic repository for nuclear waste, essentially ending nearly 30 years of effort at great expense to develop a strategy for the final disposal of nuclear waste created by commercial nuclear power plants and the defense programs that began during WWII and continued through the Cold War. Most recently, there has been a nuclear “renaissance” with increased interest and expectations for the role of nuclear power as a CO2-free source of energy. New proposals for nuclear power production include, not only an increase in the number of nuclear power plants, but also complicated schemes for reprocessing spent nuclear fuel, and in some cases, transmutation of some nuclear waste. In this presentation, I will compare the present scale of the nuclear waste problem (i.e., sources, volumes and activities) in the United States against the progress that has been made during the past 30 years for management and/or disposal. I will summarize my view of the principal issues that have prevented the United States from arriving at a strategy for implementing a solution for the disposal of SNF and HLW. Finally, I will discuss the lessons that must be learned, if we do not want to find the nation in the same position 30 years from now (also see R.C. Ewing and Frank N. von Hippel, 2009, Science, 235, 151-151).
12:30 PM - **AA1.7
The Role of the Actinides in the Performance Assessment of a Nuclear Waste Repository. SKB’s Supporting Actinide Research.
Lars Werme 1 , Sergei Butorin 1 , Peter Oppeneer 1
1 Department of Physics and Materials Science, Uppsala University, Uppsala Sweden
Show AbstractAfter a few hundred years, the actinides will dominate the radiotoxicity of spent nuclear fuel. This does not necessarily mean that the actinides will dominate the dose to organisms at the surface above a geologic repository. Quite the contrary, in most performance assessments the dose is dominated by long-lived fission products, activation products and, in the very long perspective, actinide daughters. This makes the far-field migration properties of the actinides less interesting for further research. There are, however, other aspects of the presence of actinides in spent nuclear fuel that require further attention. With increasing fuel burnup, the content of higher actinides increases in the fuel and the actinides have their highest concentration at the periphery of the fuel pellet. This leads to an increase in alpha activity at the fuel surface and an increased fission rate, i.e., a higher burnup and also a re-crystallization of the rim of fuel pellet. The chemical stability in water of this re-structured material needs to be addressed. The increased alpha activity also results in a helium build-up in the uranium dioxide fuel matrix. This may have consequences for the stability and the mechanical integrity of the fuel matrix.Evaluation of the consequences of the alpha surface dose rate and water radiolysis for possible fuel oxidation requires knowledge of the redox properties of the actinides and their possible oxidation states at the fuel surface. The possibilities of electron transfer to oxidized actinide species as well as possible electron donors in the vicinity of the fuel also require attention. In the longer perspective when the alpha activity has decayed, chemical dissolution of the fuel matrix can occur and then the “solubility” of the uranium dioxide fuel matrix will be important for the performance assessment. These issues and SKB sponsored research on these will be presented and discussed.
AA2: Spent Nuclear Fuel
Session Chairs
Tuesday PM, April 06, 2010
Room 3010 (Moscone West)
2:30 PM - AA2.1
Selective Radionuclide (Cs+, Sr2+, and Ni2+) Ion-exchange by K2xSn3-xMgxS6 (x=0.5-0.95) (KMS-2).
Joshua Mertz 1 , Mercouri Kanatzidis 1 2
1 Department of Chemistry, Northwestern University, Evanston, Illinois, United States, 2 Materials Science Division , Argonne National Laboratory, Argonne, Illinois, United States
Show AbstractNew materials for the removal of radioactive waste streams from nuclear power plants are sorely needed to reduce waste and cost in the nuclear energy industry. 137Cs+ and 90Sr2+, both byproducts of the fission process, make up the majority of high-level waste from nuclear power plants because their daughter compounds emit high level gamma and beta particles respectively. 63Ni2+ is a byproduct of the erosion-corrosion process of the reactor components in nuclear energy plants. The concentrations of these ions in solution determine the Waste Class (A,B, or C) and thus selective removal of these ions over large excesses of other ions is necessary to reduce waste and cut costs. Herein we report the use of the Inorganic Ion Specific Media (ISM) K2xSn3-xMgxS6 (x=0.5-0.95) (KMS-2) for the ion exchange of Cs+, Sr2+, and Ni2+ in several different conditions. We will also report the stability of this new material in the general conditions found at nuclear power plants (pH 6-8) and DOE sites (pH >10). Measurements at low concentrations were followed by inductively coupled plasma mass spectrometry and Kd values are reported for each of the ions in a variety of conditions.
2:45 PM - AA2.2
Drawdown of Actinide Chlorides from Electrorefiner Salt via Lithium Reduction.
Michael Simpson 1 , Daniel LaBrier 2 , Michael Lineberry 2 , Tae-Sic Yoo 1
1 Pyroprocessing Technology, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Idaho State University, Idaho Falls, Idaho, United States
Show Abstract Electrorefining of spent nuclear fuel can be used to separate U, Pu, and minor actinides from fission products and other fuel constituents for eventual fabrication of fast reactor fuel. In this process, fission products as well as transuranic elements become oxidized to chloride form and accumulate in the molten salt electrolyte. When the extent of salt contamination by fission products reaches a pre-determined limit, the salt must be removed and either discarded or treated to remove fission products and returned to the electrorefiner. In either case, fission products are diverted into a waste stream. To prevent actinides from being carried along with the fission products, it is necessary to perform a drawdown operation on the salt to remove actinides from the salt phase. Various methods have been previously investigated to achieve this drawdown, including reactive extraction into a molten metal phase and electrolysis onto a solid metal cathode. While these approaches appear to be feasible, there are operational complexities involved with them. A new process has been proposed by researchers at Idaho National Laboratory involving reaction with lithium or another active metal that offers the potential to be significantly simpler than these other methods. After such a drawdown operation, the actinide-free salt can be transferred to the waste process equipment via draining or pumping. This can be followed by re-chlorination of the actinides and return of the actinides to the electrorefiner. While simple in principal, a key technical issue pertaining to this approach is the ability to reduce all of the actinides while minimizing reduction of rare earth fission products. If complete removal of actinides necessarily causes significant rare earth reduction, it will prove to be difficult to get rare earths out of the process salt and into the waste stream. Early experiments have been completed and will be reported involving only rare earth elements that were used to study the overlap of drawdown between different species based on free energy of formation. It has been experimentally shown that the ability to achieve selective drawdown is highly dependent upon the free energy gap. Impact of this observation on the ultimate goal of performing complete drawdown of actinides will be discussed.
3:00 PM - AA2.3
Cold Crucible Vitrification of U-bearing SRS SB4 HLW Surrogate.
Sergey Stefanovsky 1 , Alexander Ptashkin 1 , Oleg Knyazev 1 , Olga Stefanovsky 1 , James Marra 2
1 , SIA Radon, Moscow Russian Federation, 2 , Savannah River National Laboratory, Aiken, South Carolina, United States
Show AbstractThe material at 55 wt.% Sludge Batch 4 (SB4) high level waste (HLW) loading was produced in the demountable 56 mm inner diameter cold crucible and spontaneously cooled to room temperature in the cold crucible and in alumina crucible in a resistive furnace by a canister centerline cooling (CCC) regime. In total, a batch mixture in amount of ~900 g was fed to the CCIM and a glass in amount of ~660 g was produced. Average melt production rate, specific melt production rate and melting ratio under steady-state conditions were 0.16 kg/hr, 14.5 kg/(dm2d) and 33.1 kw hr/kg, respectively. XRD patterns of the materials sampled from the upper, middle and lower (near-bottom) parts of the block and cooled by the CCC regime demonstrate their similarity and are composed of major vitreous phase and minor spinel structure phase with d-spacing parameters close to magnetite/trevorite (Fe,Ni)Fe2O4 solid solution. The spinel phase is different in various parts of the block produced in the cold crucible and cooled by the CCC regime. No separate U-bearing phases were found.
3:15 PM - AA2.4
Capture and Sequestration of Radioactive Iodine.
Brian Westphal 1 , Ken Bateman 1 , Dennis Wahlquist 1 , William McCartin 1 , Jang-Jin Park 2 , Jin-Myeong Shin 2 , Bruce Begg 3
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Korea Atomic Energy Research Institute, Daejeon Korea (the Republic of), 3 , ANSTO Inc., Idaho Falls, Idaho, United States
Show AbstractConsidering the toxicity and mobility of radioactive iodine, its capture and sequestration is important following the processing of spent oxide fuel. Whether the process flowsheet for spent oxide fuel contains aqueous or pyrometallurgical methods, complete iodine capture can be achieved upfront as a head-end operation for both options. If a high temperature (>1000oC) oxidative head-end step is included in the flowsheet, iodine can be entirely volatized and collected on filter media by chemical adsorption.Trapping experiments have been performed at the Idaho National Laboratory (INL) to assess the performance of AgX sorbent media during the oxidation of spent LWR oxide fuel. Since the emphasis of the oxidation step at the INL has been fuel decladding, the process has been termed DEOX. Demonstration of complete iodine release from the spent fuel and capture has been accomplished with laboratory-scale equipment in a hot cell environment [1]. The maximum performance of the AgX media has been 75 ug iodine/g media/g fuel processed which compares favorably to other research with irradiated fuel [2]. In addition to iodine, significant quantities of tritium have also been collected on the AgX filter media. Testing is ongoing to increase the iodine loading and efficiency. Based on the encapsulation of surrogate iodine-bearing sorbent media [3], AgX media loaded with radioactive iodine from DEOX testing has been sequestered in a tin matrix by hot isostatic pressing (HIP). The placement and containment of the iodine sorbent media was examined by neutron radiography following the HIP cycle and confirmed the successful sequestration of the iodine. Additional destructive analyses are pending.References[1] B.R. Westphal, J.J. Park, J.M. Shin, G.I. Park, K.J. Bateman, and D.L. Wahlquist, “Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System”, Sep. Sci. Tech., 43, 9-10 (2008), pp. 2695.[2] H. Mineo, M. Gotoh, M. Iizuka, S. Fujisaki, H. Hagiya, and G. Uchiyama, “Applicability of a Model Predicting Iodine-129 Profile in a Silver Nitrate Silica-Gel Column for Dissolver Off-Gas Treatment of Fuel Reprocessing”, Sep. Sci. Tech., 38, 9 (2003), pp. 1981.[3] E.R. Vance, D.S. Perera, S. Moricca, Z. Aly, and B.D. Begg, “Immobilization of 129I by Encapsulation in Tin by Hot-Pressing at 200oC”, J. Nucl. Mat., 341 (2005), pp. 93.
3:30 PM - **AA2.5
Natural and Experimental Studies of Uranium Sequestration by TiO2.
Mostafa Fayek 1 , Ren Zhang 1 , Feiyue Wang 1
1 Geological Sciences, University of Manitoba, Winnipeg, Manitoba, Canada
Show AbstractThe nuclear fuel cycle involves a number of steps including mining, enrichment, nuclear power generation, and disposal of spent nuclear fuel. While nuclear energy provides one of few promising, cleaner alternatives to fossil fuel, its development and acceptance have been challenged by the high initial cost of construction of nuclear power plants and concerns over the environmental and human health impact associated with mine tailings and disposal of high-level nuclear waste (HLNW). Technological developments have reduced construction costs and improved nuclear power plant safety; however, the environmental and human health impact related to U mine tailings and disposal concepts (e.g., geological repositories) for HLNW is still hotly debated. Therefore, developing techniques to sequester U that may be released to the environment from mine tailings or HLNW repositories would improve public confidence in the entire fuel cycle. Uranium mineralization at the Nopal I U deposit, Peña Blanca District, Mexico is exposed at surface and extends downwards ~100 meters and stops ~130 meters above the water table. Exposure of the U ore to meteoric water has caused significant oxidation and remobilization of U in the vicinity of the deposit. However, U concentrations in the groundwater are <50 ppb. In this regards, the Nopal U deposit is an excellent natural laboratory to study U mobilization in near surface environments. Measurement of α, β, γ radiation profiles in a continuous drill core through the deposit found anomalous radiation (~1000 cps) ~ 90 meters below the deposit and 40 meters above the water table. At this depth the rock unit is a highly altered conglomerate that is clay-rich with large fractures infilled with clay. Back-scattered electron (BSE) imaging, elemental mapping, scanning transmission electron microscope (STEM) images, and electron diffraction patterns show that this rock unit consists of disseminated grains of anatase as well as anatase replacing highly altered sphene. Uraninite is intimately associated with the anatase, which suggests that anatase sequestered remobilized U before it reached the water table. Based on the strong association of U with anatase in natural uraniferous systems, batch experiments show that sorption of aqueous U6+ on TiO2 occurs rapidly under oxic conditions in a wide pH range (2-11) at room temperature. XPS analysis of U-sorbed TiO2 showed that both U6+ and U4+ exist and U4+ is the dominant species. HRTEM study of U-sorbed anatase showed that U is enriched in an outer amorphous layer and occurs in crystalline form. Selected area diffraction shows that the crystalline form is UO2 whereas U in the amorphous layers is likely U6+. The results suggest that aqueous U6+ can be effectively removed by TiO2 through both sorption of U6+ and precipitation of reduced U4+, indicating that TiO2 can be used as a novel, cost-effective U getter for U mine tailing sites, contaminated aquifers, and HLNW containers.
Symposium Organizers
Katherine L. Smith Embassy of Australia
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Scott Kroeker University of Manitoba
Blas Uberuaga Los Alamos National Laboratory
AA3: Glass Wasteforms
Session Chairs
Wednesday AM, April 07, 2010
Room 3010 (Moscone West)
9:30 AM - **AA3.1
Structural Evolution of Nuclear Glasses under Forcing Conditions (Alteration, Irradiation).
Georges Calas 1 , Laurence Galoisy 1 , Laurent Cormier 1 , Sylvain Peuget 2 , Jean-Marc Delaye 2 , Patrick Jolllvet 3
1 Mineralogy, University of Paris, Paris France, 2 DEN/DTCD/SECM/LMPA, CEA Valrhô-Marcoule, 30207 Bagnols-sur-Cèze cedex France, 3 DEN/DTCD/SECM/LCLT, CEA Valrhô-Marcoule, 30207 Bagnols-sur-Cèze cedex France
Show AbstractAssessing the long-term behavior of nuclear glass implies the prediction of their long-term performance, and more precisely of their evolution with irradiation and during interaction with water. After briefly recalling the major characteristics of the local and medium-range structure of borosilicate glasses of nuclear interest, we will present some structural features of this evolution under forcing conditions. Specific structural tools used to determine the local structure of glass surfaces include synchrotron-radiation x-ray absorption spectroscopy with total electron yield detection. The evolution of the structure of glass surface has been determined in irradiated (B, Zr) and altered (Zr, Fe, Si, Al) glasses. During alteration in near- or under-saturated conditions, some elements such as Fe change coordination, as other elements such as Zr only suffer structural modifications in under-saturated conditions. These structural modifications may explain the chemical dependence of the initial alteration rate and the transition to the residual regime. They also illustrate the molecular-scale origin of the processes at the origin of the glass-to-gel transformation. During external irradiation, there is direct evidence of a coordination change of B at the glass surface. In addition, for a better understanding of the modification of glass structure by heavy ions, complementary information is provided by molecular dynamics simulations, showing a combination of elastic and inelastic effects. These two processes are suspected to produce a modification of different physical properties of nuclear glasses during irradiation.
10:00 AM - AA3.2
Structural and Crystallization Study of a Simplified Aluminoborosilicate Nuclear Glass Containing Rare-earths: Effect of ZrO2 Concentration.
Daniel Caurant 1 , Arnaud Quintas 1 , Odile Majerus 1 , Thibault Charpentier 2 , Pascal Loiseau 1 , Dominique de Ligny 3 , Jean-Luc Dussossoy 4
1 Laboratoire de Chimie de la Matière Condensée de Paris (UMR 7574) Chimie-ParisTech (ENSCP), CNRS, Paris France, 2 IRAMIS Service Interdisciplinaire sur les Systèmes Moléculaires et Matériaux , CEA Saclay, Gif -sur-Yvette France, 3 Laboratoire de Physico-Chimie des Matériaux Luminescents (UMR 5620) Université Claude Bernard Lyon1, CNRS, Villeurbanne France, 4 Laboratoire d’étude et développement de matrices de conditionnement DEN, MAR/DTCD/SECM, CEA Marcoule, Bagnols-sur-Cèze France
Show AbstractZrO2 is introduced in glass compositions for different applications. For instance, it is known to act as nucleating agent in glass-ceramics and to increase glass alkali-resistance in cement. ZrO2 is also present in borosilicate glass compositions used to immobilize highly radioactive nuclear wastes. In this case, zirconium may originate both from the highly radioactive waste solutions (arising from nuclear spent fuel reprocessing) and from the glass frit added to the wastes for glass preparation.In this paper, we present the study of the effect on glass structure and crystallization tendency of increasing ZrO2 concentration (from 0 to 6 mol%) in a simplified new nuclear glass composition belonging to the SiO2-Al2O3-B2O3-Na2O-CaO-ZrO2-RE2O3 system (with RE= Nd or La) developed to incorporate rare earth-rich wastes. The introduction of ZrO2 induced an increase of the glass transformation temperature and of the compacity of the oxygen network. The structural evolution of the glassy network was followed by 27Al, 29Si, 23Na, 11B MAS NMR and Raman spectroscopy whereas the environment of Nd3+ cations was followed by optical absorption and EXAFS spectroscopies. The environment of Zr4+ cations was also probed by EXAFS. Whereas a decrease of the proportion of BO4 units was observed, only a small effect occurred on the environment of AlO4 units. Nevertheless, according to Raman spectra, a significant structural evolution of the silicate network seems to occur when [ZrO2] increased. The crystallization tendency of the supercooled melt was studied either during slow cooling (1°C/min, i.e. close to the natural cooling rate in the bulk of current borosilicate nuclear glass containers after casting) or after nucleation + crystal growth thermal treatments. For all samples, the crystallization of only a rare-earth silicate apatite phase was observed. Whereas nucleation mainly occurred from the surface of the samples without ZrO2, the introduction of zirconium induced increasing apatite crystallization in their bulk showing the nucleating effect of ZrO2 for the composition studied in this work.
10:15 AM - AA3.3
Precipitation of Mixed-alkali Molybdates During HLLW Vitrification.
Scott Kroeker 1 , Carolyn Higman 1 , Vladimir Michaelis 1 , Nicholas Svenda 1 , Sophie Schuller 2
1 Chemistry, University of Manitoba, Winnipeg, Manitoba, Canada, 2 DEN/DTCD/SECM/LDMC, CEA Valrhô Marcoule, Bagnols/Céze France
Show AbstractCrystalline precipitates from molybdenum-containing nuclear waste glasses are complex, often containing multiple cations which confound routine structural techniques. A simplified mixed-alkali borosilicate model glass was found to have minor crystalline phases which could not be identified by x-ray diffraction. Multinuclear magnetic resonance (NMR) spectroscopy revealed sharp peaks characteristic of crystallinity superimposed on the broader glass signals, but were unattributable to any known molybdate phases. When a comprehensive range of cesium molybdates failed to reveal any matches with the observed 133Cs magic-angle spinning (MAS) NMR peaks in the composite glass/crystalline material, a series of mixed-alkali sodium-cesium molybdate phases was synthesized. 23Na, 133Cs and 95Mo MAS NMR enabled the assignment of x-ray diffraction powder patterns of the complex phase assemblages, revealing the formation of several mixed-cation molybdates which correlated with the observed NMR peaks for the phase-separated model glass. This work highlights the prominence of multiple crystalline phases in molybdenum-bearing nuclear waste glasses, and demonstrates the unique utility of solid-state NMR as a fingerprinting approach to identifying complex phases, especially where x-ray diffraction is limited by multiple phases or substitutionally disordered precipitates.
10:30 AM - AA3.4
Study of the Pd-Te-Ru System in Sodium Borosilicate Waste Glasses.
Stephane Gosse 1 , Sophie Schuller 2 , Christine Gueneau 1
1 Physico-Chemistry Department, Commissariat à l'Energie Atomique, Gif-sur-Yvette France, 2 Department of Waste Treatment and Conditioning, Commissariat à l'Energie Atomique, Bagnols-sur-Cèze France
Show AbstractThe platinoid elements (Pd, Ru, Rh) of very low solubility in high level radioactive borosilicate glasses precipitate both under (Pd-Te, Ru-Rh, Ru) metallic particles [1] and (RuO2, RhO2) oxide phases with acicular or polyhedral shapes [2] during the vitrification process. Composition and microstructure evolutions of these phases can affect significantly the physico-chemical properties of the melt such as viscosity, electrical conductivity and thermal conductivity during melting in an induction melting cold crucible.Several studies are undertaken at CEA [1-2] in order to point out the reactions and the chemical interactions in the liquid and viscous states between the glass matrix and the platinoids issuing from the calcinated waste.Among these studies, a thermodynamic database is being developed on the metallic (Pd-Rh-Ru-Te) and oxide (O-Pd-Rh-Ru-Te) systems. In this work based on the CALPHAD method, the Gibbs energies of each phase is modelled in order to provide an overall thermodynamic description of the platinoid phases in nuclear waste glasses.The main objective of the database is to calculate phase diagrams and thermodynamic properties. Also, this flexible tool enables to justify the relative stability between metallic and oxide phase in function of both the temperature and the oxygen potential fixed by the glass frit.At this point, the (Pd-Te, Pd-Ru, Ru-Te) binary sub-systems have been modelled. The calculations have been compared with experimental thermodynamic data from literature. Then, the Pd-Te-Ru ternary system built by extrapolation of the binaries enables to calculate isothermal cross-sections and thermodynamic properties in the 773 K-1523 K temperature range so as to characterise the behaviour of the metallic platinoid phases in waste glasses.Some solidification routes are also calculated for palladium and tellurium compositions corresponding to those analysed in the glasses. They enable to predict the composition of the Pd-Ru-Te phases at the thermodynamic equilibrium as well as an estimate of the solubility limit of the ruthenium in Pd-Te alloys in compliance with experimental results.[1] Structure of Pd-Te precipitates in a simulated high-level nuclear waste glassL. Galoisy, G. Calas, G. Morin, S. Pugnet, C. Fillet, 1998J. Mater. Research, Vol. 13, N°. 5[2] Behaviour of ruthenium dioxide particles in borosilicate glasses and melts Rachel Pflieger, Leila Lefebvre, Mohammed Malki, Mathieu Allix, Agnès Grandjean, 2009J. Nuclear Mater, Vol. 389, N°3
10:45 AM - AA3.5
The Effect of Increased Waste Loading on the Durability of High Level Waste Glass.
Chris Brookes 1 , Mike Harrison 2 , Andrew Riley 1 , Carl Steele 1
1 High Level Waste Plants, Sellafield Ltd, Seascale, Cumbria, United Kingdom, 2 , National Nuclear Laboratory, Seascale, Cumbria, United Kingdom
Show AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors arising from the reprocessing of spent nuclear fuel within glass to provide a stable and durable waste form suitable for safe long term storage and ultimate disposal.WVP processes liquors from the reprocessing of both Magnox and Oxide spent nuclear fuel. Magnox feed is relatively low in fission products but contains significant amounts of Al and Mg from the fuel cladding. Oxide feed, from LWR and AGR spent fuel, is of higher burnup and contains more fission products, along with Gd and other process additives. Oxide feed is mixed with Magnox waste in order to yield a Blend product. The target waste oxide incorporation rate for both Blend and Magnox glasses is 25 wt%. However, recent programmes have established WVP operational envelopes for increased waste loading.Currently, work is progressing on understanding the durability of WVP Product Glass to underpin its suitability for deep geological disposal, the U.K.’s preferred disposal route for HLW and ILW. This paper describes the results from static leach tests using the ASTM standard MCC-1 and PCT protocols that were performed on inactive HLW glasses fabricated at full scale on the Sellafield Vitrification Test Rig. The samples comprised monoliths and powders of a 75:25 Oxide:Magnox Blend glass with 31 wt% waste incorporation and a Magnox-only glass with 35 wt% waste incorporation. The tests were carried out in de-ionised water at 90 °C for durations of up to 42 days and normalised mass losses calculated.The results of MCC-1 and PCT tests on both 31 wt% Blend and 35 wt% Magnox glasses, showing measurable differences to the corresponding standard 25 wt% waste incorporation glasses, are presented. A series of SEM investigations were also undertaken, enabling the surface of the leached glass samples to be studied without disturbing alteration layers formed during the tests. The variation in composition and thickness of the alteration layer with sample type and duration is reported.
AA4: Cementitous Wasteforms
Session Chairs
Wednesday PM, April 07, 2010
Room 3010 (Moscone West)
12:00 PM - AA4.2
Radioactive Iodine Separations and Waste Forms Development.
Tina Nenoff 1 , James Krumhansl 2 , Terry Garino 3
1 Surface and Interface Sciences, Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 Geochemistry Department, Sandia National Laboratories, Albuquerque, New Mexico, United States, 3 Electronic & Nanostructured Materials Department, Sandia National Laboratories, Albuquerque, New Mexico, United States
Show AbstractReprocessing nuclear fuel releases gaseous radio-iodine containing compounds which must be captured and stored for prolonged periods. Ag-loaded mordenites are the leading candidate for scavenging both organic and inorganic radioiodine containing compounds directly from reprocessing off gases. Alternately, the principal off-gas contaminant, I2, and I-containing acids HI, HIO3, etc. may be scavenged using caustic soda solutions, which are then treated with bismuth to put the iodine into an insoluble form. Our program is focused on using state-of-the-art materials science technologies to develop materials with high loadings of iodine, plus high long-term mechanical and thermal stability. In particular, we present results from research into two materials areas: (1) zeolite-based separations and glass encapsulation, and (2) in-situ precipitation of Bi-I-O waste forms. Ag-loaded mordenite is either commercially available or can be prepared via a simple Ag+ ion exchange process. Research using an Ag+-loaded Mordenite zeolite (MOR, LZM-5 supplied by UOP Corp.) has revealed that I2 is scavenged in one of three forms, as micron-sized AgI particles, as molecular (AgI)x clusters in the zeolite pores and as elemental I2 vapor. It was found that only a portion of the sorbed iodine is retained after heating at 95o C for three months. Furthermore, we show that even when the Ag-MOR is saturated with I2 vapor only roughly half of the silver reacted to form stable AgI compounds. However, the Iodine can be further retained if the AgI-MOR is then encapsulated into a low temperature glass binder. Follow-on studies are now focused on the sorption and waste form development of Iodine from more complex streams including organo-iodine compounds (CH3I). Bismuth-Iodate layered phases have been prepared from caustic waste stream simulant solutions. They serve as a low cost alternative to ceramics waste forms. Novel compounds have been synthesized and solubility studies have been completed using competing groundwater anions (HCO3-, Cl- and SO42-). Distinct variations in solubility were found that related to the structures of the materials.Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the US DOE’s NNSA under contract DE-AC04-94AL85000.
12:15 PM - AA4.3
Immobilization Mechanisms of Dissolved Ionic Species in Cement Matrix.
Mostafa Youssef 1 , Bilge Yildiz 1
1 Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractA major scenario in which high-level radioactive waste disposed in a geological repository might reach the biosphere is groundwater seeping into the repository followed by the corrosion of waste canisters and leaching of radionuclides into groundwater. For this pathway, we consider cementitious matrices as the waste containment medium. It is important to fundamentally understand the immobilization mechanism and binding capacity of radionuclides in the cement matrix . For this purpose, we focus on the fission products Strontium-90 and Cesium-137, each with half-life of about 30 years, accounting for the bulk of the radioactivity and decay heat in spent nuclear fuel for several decades after a few years upon discharge from the nuclear reactor. Our approach utilizes atomistic simulation techniques. We adopted a recently developed molecular model for the highly disordered gel Calcium-silicate-Hydrate (C-S-H), the most abundant and main binding phase in cement, as our working prototype. This solid phase is known to be particularly relevant for the uptake of metal cations. The molecular model to be used successfully accounts for several structural, mechanical and chemical properties of this complex material without being fitted to reproduce them. In our simulations, we incorporate a set of interatomic potentials that was used intensively to study the interaction between the surfaces of inorganic solids and aqueous solutions.The first step in this research is to examine the possibility of cationic exchange between the two radionuclides and Ca2+/Si4+ on the surface of C-S-H by means of energy-minimization. The effect of ion exchange on the structural integrity of C-S-H is also considered. This is followed by creating a C-S-H/ water interface in the presence of radionuclides ions in the form of dissolved salts and simulating the structural evolution at finite temperature. Such an interface between water and C-S-H is naturally omnipresent in the cement matrix. Identification of the type of surface complexation (inner or outer) and the residence times of these ions in their complex spheres is of importance to evaluate how tightly they can be bound to C-S-H surface. Our quantitative considerations of the chemical processes involved in the immobilization of the radionuclides have a direct impact in the assessment of the long-term performance of cementitious wasteforms.
12:30 PM - AA4.4
Assessment of Concrete Cracking at Nuclear Waste Disposal Facilities via Fiber Optic Sensors.
Sanaan Lair 1 , Antonio Motta 1 , John Walton 1 , Arturo Woocay 1
1 Civil Engineering, University of Texas at El Paso, El Paso, Texas, United States
Show AbstractEngineered concrete barriers used for the purpose of radioactive waste disposal must maintain their integrity for long periods of time in order to prevent the release of radionuclides into the environment. Degradation of the concrete can lead to failure of the engineered barrier to isolate the radioactive waste from the surrounding environment. One of the main degradation methods of concrete is crack formation. Cracks can be formed in concrete by multiple processes including physical loading, drying shrinkage, reinforcement corrosion, thermal stresses, subgrade settlement, physical loading and carbonation. Cracks form preferential pathways for fluid flow and mass transport. Federal regulations require that performance assessments of radioactive waste disposal facilities be conducted in order to prove technical compliance with the regulatory standards and to demonstrate that the facility will achieve the stated performance objectives. As part of the performance assessment, assumptions are made about initial cracking and crack formation over time to predict the useful life of the facility. These assumptions must be verified by monitoring the structure and comparing actual results to the assumptions. Current methods of crack detection consist mainly of visual inspection which is inaccurate and not suited to buried concrete vaults. Other nondestructive test methods which are widely used usually do not detect very small cracks and are unable to determine crack width. Fiber optic sensors offer a novel approach to crack monitoring and offer the possibility of determining the amount, width and location of the cracks as they form without any prior knowledge of where these cracks will form. A distributed system of fiber optic sensors may be embedded in the concrete structure; the formation of cracks causes the fiber to bend and a change in the signal indicates the location and size of the crack. Crack formation can change the air permeability of the concrete structure, therefore rapid fluctuations in air pressure may indicate the presence of cracks as they allow variations in barometric pressure to propagate into the structure. When cracks form they can fill with water which has a high heat capacity; therefore cracks may also be monitored by observing temperature variations. Fiber optics can also be employed to monitor pressure and temperature changes inside the vault, at the surface and in the surrounding soil to indicate the presence of cracks over time. This paper explores the use of fiber optic sensors in monitoring concrete degradation of nuclear waste disposal facilities and comparing the results to assumptions made during performance assessments to measured cracking during the first ~20 years of actual performance. Careful consideration is given to false positive and false negative signals in the monitoring systems through the use of multiple independent methods of measurement.
12:45 PM - AA4.5
New Lanthanide or Uranium Oxalato-nitrates Crystallized From Acidic Solutions.
Christelle Tamain 2 , Murielle Rivenet 1 , Benedicte Chapelet-Arab 2 , Stephane Grandjean 2 , Francis Abraham 1
2 CEA VALRHO Centre de Marcoule, DEN/DRCP/SCPS/Laboratoire de Chimie de Actinides, Bagnols sur Cèze France, 1 ENSCL/USTL, Unité de Catalyse et de Chimie du Solide, Villeneuve d'Ascq France
Show AbstractOxalic acid is a very common reagent to recover actinides from radioactive liquid waste using precipitation methods because of the very low solubility of An(IV) or An(III) oxalate compounds in acidic solutions. The oxalic precipitation of plutonium is widely used at an industrial scale during the reprocessing of the nuclear fuel, e.g. the PUREX process converts this energetically valuable actinide into oxide. Recently our group showed that the flexibility of the oxalate ligand allowed the formation of mixed An(IV)–An(III) actinides oxalate solid compounds based on two or three-dimensional actinide-oxalate frameworks. As these materials are particularly suitable precursors of actinides oxide solid solutions, the actinides co-precipitation is one option for the co-management of actinides in an integrated closed fuel cycle currently under evaluation for Generation III/IV systems. In these oxalates, An(III) and An(IV) occupy the same crystallographic site, the charge compensation being insured by monovalent ions such as hydrazinium ions which are present in the acidic solutions to prevent the oxidation of U(IV) in presence of nitrate ions.In some conditions, the incorporation of nitrate species in the solids cannot be ruled out. To identify the various oxalato-nitrates likely to form we studied the crystallization of such compounds by various methods (diffusion, hydrothermal syntheses...) in different conditions in presence of hydrazinium ions. In the first stage, lanthanides were used as surrogates of the actinides (III) radioactive elements. Single crystals of different compounds were grown corresponding to various oxalate/Ln(III) ratio and containing nitrates as bidentate ligands or as counter ions. In most compounds hydrazinium ions are present as counter ions. Uranium compounds were also investigated. This communication reviews the various oxalato-nitrates of lanthanide or uranium obtained by crystallization from nitric acid solution containing hydrazinium ions. The crystal growth methods are described and the crystal structures, determined by X-ray diffraction from single crystals, are discussed in terms of metal-oxalate frameworks. For example, the adjustment of the conditions of diffusion of ions allowed us to synthesize lanthanide (III) oxalates with structure based on a neutral three dimensional lanthanide (III) arrangement [Ln2(C2O4)3(H2O)3] creating cavities occupied by both negative (NO3)- and positive (N2H5)+ ions. In some cases, hydrazinium is present as (N2H6)2+ ions.
AA5/Z7: Joint Session: Actinide Chemistry
Session Chairs
Thomas Fanghaenel
Lynne Soderholm
Blas Uberuaga
Wednesday PM, April 07, 2010
Room 3008 (Moscone West)
2:30 PM - **AA5.1/*Z7.1
High-resolution 17O NMR Nuclear Magnetic Resonance Studies of Uranium Oxides: Preliminary Results.
Ian Farnan 1 , Kevin Boland 2 , David Clark 3
1 Earth Sciences, University of Cambridge, Cambridge United Kingdom, 2 Inorganic, Isotope, and Actinide Chemistry (C-IIAC, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 Seaborg Institute, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractNuclear magnetic resonance is an element specific technique that can provide unique quantitative information about atomic distributions amongst different sites in a material. In the case of actinides, the large onsite hyperfine coupling of 5f electrons to the nucleus means that nuclear spin relaxation lifetime effects will make direct observation of the actinide nucleus extremely difficult. However, observation of resolved transferred hyperfine shifts at adjacent nuclei may well be an extremely powerful probe of local sites and their occupation. We have applied magic-angle spinning NMR to U17O2 to determine the resolution possible as a proof of principle experiment. We have obtained a static NMR spectrum of U17O2 at 9.4 Tesla with a width of 42 kHz (FWHM) in agreement with literature values for static spectra. Subsequent MASNMR spectra of the same sample with increasing spinning speeds of 5, 10, 15 kHz show that the broad line shape breaks up into a central band and sidebands such that resolution increases with increased spinning speed. A minimum centreband linewidth at 15 kHz spinning of 3.2 kHz is obtained. This represents a width of 60 ppm in terms of local field, thus resolution of transferred hyperfine shifts differing by 6-10 ppm should be possible. The total NMR shift observed for U17O2 is 726 ppm from H2O. Temperature dependent measurements of the shift indicate that the Fermi contact contribution is ~ 160 ppm. This indicates very little delocalisation of the U4+ 5f2 unpaired electron density to the nearby (2.35 Å) oxygen atoms. A preliminary spectrum of U4O9, which will contain adventitious oxygens, is less well-resolved at 15 kHz spinning, but indicates the presence of more than one site. Ongoing NMR resolution enhancement protocols and/or faster spinning and lower magnetic fields should make resolution and identification of oxygen sites in this material a tractable problem.
3:15 PM - AA5.3/Z7.3
Radiation Range and Damage Assessment in UO2 Simulated byClassical Molecular Dynamics.
Byoungseon Jeon 1 , Anurag Chaudhry 1 , Mark Asta 2 , Steve Valone 3 , Niels Gronbech-Jensen 1
1 Dept. of Applied Science, University of California, Davis, Davis, California, United States, 2 Dept. of materials science and engineering, University of California, Berkeley, Berkeley, California, United States, 3 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractMolecular Dynamics (MD) has been used extensively to study the crystal damage production and short time evolution in UO2 due to Primary Knock-on Atoms (PKAs). We here present an approach based on a combination of MD strategies. First, to characterize the radiation range, REED-MD [1,2] and binary collision methods [3] are used and compared with experiments on single/poly crystalline UO2. Contributions to the atomic force fields are nuclear-nuclear, electron muffin-tin drag forces, and electron stopping. The effect of the target material structure and channeling is discussed. Secondly, full massively parallel MD cascade simulations have been done to evaluate the damage on UO2 matrix, yielding displacement cascades with. Confirming the migration of defects and recovery of matrix, temporal variation of energy landscape are shown. Through extensive analysis, the behavior of defect and damage evolution will be addressed.[1] K. M. Beardmore, N. Gronbech-Jensen, Physical Review E, v.57, pp.7278-7287, 1998[2] B. Jeon and N. Gronbech-Jensen, Computer Physics Communications, v.180, pp.231-237, 2009[3] www.srim.org
3:30 PM - AA5.4/Z7.4
Electronic Structure and Ionicity of Actinide Oxides from First Principles.
Leon Petit 1 2 , Axel Svane 2 , Walter Temmerman 1 , Zdzislawa Szotek 2 , George Stocks 3
1 Computational Science and Engineering Department, Daresbury Laboratory, Warrington United Kingdom, 2 Department of Physics and Astronomy, Aarhus University, Aarhus Denmark, 3 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe ground state electronic structures of the actinide oxides AO, A2O3 and AO2 (A=U, Np, Pu, Am, Cm, Bk, Cf) are determined from first-principles calculations, using the self-interaction corrected local spin-density (SIC-LSD) approximation. Emphasis is put on the degree of f-electron localization, which for AO2 and A2O3is found to follow the stoichiometry, namely corresponding to A4+ ions in the dioxide and A3+ ions in the sesquioxides. In contrast, the A2+ ionic configuration is not favorable in the monoxides, which therefore become metallic. The energetics of the oxidation and reduction of the actinide dioxides is discussed, and it is found that the dioxide is the most stable oxide for the actinides from Np onwards. Our study reveals a strong link between preferred oxidation numberand degree of localization which is confirmed by comparing to the ground state configurations of the correspondinglanthanide oxides. The ionic nature of the actinide oxides emerges from the fact that only those compounds will form where the calculated ground state valency agrees with the nominal valency expected from a simple charge counting.
4:15 PM - **AA5.5/*Z7.5
Actinide Solid/Solution Interface Chemistry Relevant to Nuclear Waste Disposal.
Horst Geckeis 1
1 Institute of Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe Germany
Show AbstractAssessment of environmental actinide behaviour in the environment requires fundamental insight into molecular structures of relevant actinide species. Recently, we investigated in detail colloid generation, solid/liquid interface reactions [1] and solid-solution formation of various actinides [2]. A sound understanding of such processes is required in order to allow a reliable prediction on actinide mobility or retention under nuclear waste repository conditions and in contaminated sites. Spectroscopic and classical batch type experiments and quantum chemistry calculations have been applied to obtain a consistent picture on actinide speciation and structures on mineral and colloid surfaces upon outer-sphere sorption and inner-sphere surface complex formation. Beside surface phenomena, incorporation into mineral structures appears to be a common reaction for actinides with minerals. Trivalent actinides and lanthanides have been taken as fluorescent probes to study incorporation reaction mechanisms and structural features of incorporated actinide ions. Solid-solution formation with the exchange of Ca ions vs. actinide ion has been verified as a relevant reaction with calcium carbonates (calcite, aragonite and vaterite) and apatites. Under certain conditions, however, actinides may also be integrated into hydroxides [3] and aluminosilicates. [1] H. Geckeis, Th. Rabung, J. Cont. Hydrol., 102, 2008, 187-195[2] M. Schmidt et al., Dalton Trans., 2009, 6645 – 6650[3] N. Huittinen et al., J. Coll. Interface Sci., 332, 2009, 158-164
4:45 PM - AA5.6/Z7.6
Characterization of the Penetration Mechanisms of Water into Polycrystalline UO2.
Ilaria Marchetti 1 , Fabio Belloni 1 , Paul Carbol 1 , Jerome Himbert 1 , Thomas Fanghaenel 1
1 Institute for Transuranium Elements, European Commission - Joint Research Centre, Eggenstein-Leopoldshafen Germany
Show AbstractIn the event of exposure of spent nuclear fuel to groundwater in a final repository, the mobilization of radionuclides will be affected by the modes of water attack. In particular, possible mechanisms of preferential dissolution of grain boundaries rather than matrix dissolution would cause a rapid increase of the surface area exposed to groundwater, with effects on the fraction of inventory becoming available for prompt dissolution and on the overall mechanical stability of the spent fuel. We conducted static corrosion experiments with 18O-labelled water on polycrystalline UO2 at room temperature, under monitored pH and Eh conditions. Analysis of the sample matrix was carried out by means of SEM, SIMS and high-resolution profilometry, while solution analysis for the measurement of the dissolution rate of uranium was performed by ICP-MS. SIMS depth profiling on the leached pellet showed two diffusion regimes. First, shallow depth profiling up to a depth of a few tens nm showed a short-range diffusion of 18O at high concentration (in the order of ten percent) which is compatible with Fick's lattice diffusion regime. Then, a smaller deviation from the natural 18O/(16O + 18O) isotopic ratio was measured up to a depth of 20 µm, revealing a long-range, low-concentration (in the order of a few permille, raster-averaged) diffusion regime that can be attributed to the penetration of water through grain boundaries, behaving as “high-diffusivity paths”. Fisher's, Whipple's and Levine-MacCallum's models have been used to fit the long-range experimental profiles and derive a first estimate for the grain-boundary diffusion coefficient, while the lattice diffusion coefficient has been retrieved by fitting the short-range profiles with a solution of Fick's law. Our results prove to be quite realistic, in spite of their divergence from those previously reported by other authors, whose experimental approaches however involved much higher temperatures and less direct measurement techniques. In this respect, SIMS is possibly the most powerful tool for this sort of application, as it can guarantee a direct observation of both 18O short-range diffusion – with a nanometre resolution – and water diffusion at large penetration depths. This kind of studies shows the potentiality to provide an overall frame of the corrosion/diffusion phenomena involved in the water attack on UO2, and to be extended to other polycrystalline wasteforms as well.
5:00 PM - **AA5.7/Z7.7
Nano-scale Actinide-based Clusters.
Peter Burns 1
1 Department of Civil Engineering and Geological Sciences, University of Notre Dame, Notre Dame, Indiana, United States
Show AbstractThis presentation will emphasize our current research concerning actinide-based nano-scale clusters. New clusters that will be examined include those containing pyrophosphate and oxalate ligands (more than a dozen new structures.) Emphasis will include fullerene topologies of uranyl polyhedra.
5:30 PM - AA5.8/Z7.8
Conjugates of Magnetic Nanoparticle-Actinide Chelator for Used Fuel Separation.
You Qiang 1 , Maninder Kaur 1 , Andrew Johnson 2 , Hongmei Han 1 , Jozef Kaczor 2 , Andrzej Paszczynski 2
1 1Department of Physics and Environmental Science Program, University of Idaho, Mscow, Idaho, United States, 2 Environmental Biotechnology Institute, University of Idaho, Mscow, Idaho, United States
Show AbstractThere is a significant achievement recently on nuclear fuel recycle technology based on utilizing conjugates of magnetic nanoparticle-chelator (MNP-Che) to separate the acidic nuclear aqueous waste.1-3 Based on the literature review of the progress on this magnetic separation technology, we have chosen the environmentally benign oxa-diamide chelator4 to conjugate with the core-shell MNPs for nuclear waste separation, which has the potential to make the separation process more efficient than the traditional processes using organophosphorus chelators. The oxa-diamide chelator was coupled to MNPs by reaction of an acyl chloride group on the terminal end of activated oxa-diamide with primary amines introduced on the surface of the MNPs to form a stable amide bond between the chelators and MNPs. The key issues for scaling up this application are the loading density of chelators onto the MNPs for an efficient sorption and the stability of the coated MNPs under harsh process conditions (such as highly acidic). To address these issues, different coatings and reaction chemistries were used to increase the chelator loading capacity and the MNP-Che's stability. Infrared and mass spectrometers were used to study the stability of the conjugated complex. Morphologies of the conjugated complex were characterized by transmission electron microscope; magnetic properties of MNPs and coated MNPs as well as MNP-Che complex were characterized by vibrating sample microscope. We found that the polyamine used for the conjugation process dramatically increased the density of amine groups on the MNPs, which is beneficial for the actinide sorption. The silica coating for the MNPs before attachment of chelators improves chelator loading on the MNPs by providing stable attachment surface and increasing density of hydroxyl groups, which facilitates its application in the nuclear aqueous waste separation. 1. L. Nuñez, B. A. Buchholz, M. Kaminski, S. B. Aase, N. R. Brown, G. F. Vandegrift. Separation Science and Technology 31, 1393 (1996).2. C. Gruttner, V. Bohmer, A. Casnati, J. Dozol, D.N. Reinhoudt, M. M. Reinoso-Garciae, S. Rudershausena, J. Teller, R. Ungaroc, W. Verboome, P. Wang. JMMM 293, 559-566 (2005).3. R. D. Ambashta, P. K.Wattal, S. Singh, & D. Bahadur. Separation Science and Technology 41, 925-942 (2006).4. G. X. Tian, L. F. Rao, S. J. Teat, and G. K. Liu. Chemistry-a European Journal 15, 4172 (2009).
AA6: Poster Session
Session Chairs
Thursday AM, April 08, 2010
Salon Level (Marriott)
9:00 PM - AA6.1
Biomineralization of Vivianite on the Carbon Steel Surface Attacked by the Iron Reducing Bacteria.
So Yeon Lee 2 , Hideki Yoshikawa 1 , Toshiya Matsui 2
2 World Cultural Heritage Studies, University of Tsukuba, Tsukuba, Ibaraki, Japan, 1 Geological Isolation Research and Development Directorate, Japan Atomic Evergy Agency, Tokai, Ibaraki, Japan
Show AbstractInformation of corrosion factor for metal material in the soil is important. It is lead by study on the corrosion of metal material and a long term corrosion behavior. The iron remains show the corrosion behavior of such a long term while it is buried in the soil. The data shown by the remains provide useful information in high level radioactive waste (HLW) study as overpack (carbon steel) stability in the geological disposal condition. And they also give important information for the culture heritage of conservation science. There are the two fields of the microbe influence of metal material surface. First, the study of corrosion causes by microbe (microbially influenced corrosion, MIC). Second, the study of a mineral is made by a microbe (biomineralization). The information about these fields is important to research for the influence of the microbe to metal material surface. We cultured an iron reducing bacteria in a liquid medium with carbon steel and detected Vivianite(Fe(2+)3(PO4)2 8H2O) by XRD method in this study. The results showed that the corrosion was controlled by Vivianite. For example, the excavated iron remains was controlled by Vivianite.The microbe is generally an agent promoting the corrosion of metal materials, however, the product having anticorrosion effect as the Vivianite is also generated by microbe. It is found new method of anticorrosion by using iron reducing bacteria. We static cultured the iron reducing bacteria for 41days with carbon steel. Theresult of this experiment showed that iron reducing bacteria gave the corrosion of iron material. After the incubation, we analyzed the corrosion product by XRD and SEM. The complex (biofilm, bacteria, etc) was generated by the iron reducing bacteria and that covered the carbon steel. We observed the corrosion product formed some products of green and white crystal by using microscope. They estimated needle-shaped product and lozenge crystal by SEM observation. The green crystal of Vivianite was 50∼250μm. In a corrosion process of iron material surface, iron ion Fe(2+) is dissolved from the iron metal as anode reaction, and generated to Fe(3+) oxide as a corrosion product. We considered that Vivianite is also generated as corrosion products in rich environment for Fe(2+) and phosphate by the activity of iron reducing bacteria. We got some data about morphological feature of these corrosion products.
9:00 PM - AA6.10
Effect of Ionic Strength on the Stability of Colloids Released from Injection Grout Silica Sol.
Pirkko Holtta 1 , Mari Lahtinen 1 , Martti Hakanen 1 , Jukka Lehto 1 , Piia Juhola 2
1 Department of Chemistry, Lab. of Radiochemistry, University of Helsinki, Helsinki Finland, 2 , Posiva Oy, Eurajoki Finland
Show AbstractIn Olkiluoto Finland colloidal silica called silica sol (EKA Chemicals) has been tested as a non-cementitious grout for the sealing of fractures of the hydraulic apertures of 0.05 mm or less. The use of colloidal material has to be considered in the long-term safety assessment of a spent nuclear fuel repository. The potential relevance of colloid-mediated radionuclide transport is highly dependent on their stability in different geochemical environments. Release and stability of silica colloids was followed earlier in low salinity Allard and saline OLSO reference groundwater [1, 2]. Objective of this work was to study the effect of ionic strength on stability of silica colloids released from silica gel. To use silica sol as a grout, the particles have to aggregate and form a gel within a predictable time by using sodium chloride as an accelerator. Silica gel samples were stored in contact with NaCl and CaCl2 electrolyte solutions (1 M-10-7 M) and in deionized water. Colloid release and stability were followed for two years by taking the samples after one month and then twice in a year. The release and stability of colloids were followed by measuring particle size, zeta potential, colloidal and reactive silica concentrations. The particle size distributions were determined applying the dynamic light scattering (DLS) method and zeta potential based on dynamic electrophoretic mobility. The colloidal silica concentration was calculated from DLS measurements applying a calibration using a standard series of silica sol. In 10-7–10-2 M NaCl and 10-7–10-3 M CaCl2 solutions, the mean number based colloid diameter was less than 100 nm and the colloid size distribution was rather constant. High negative zeta potential values also indicated the existence of rather stable silica colloids. After two years, the mean particle diameter was increased but it was still less than 500 nm and absolute value of zeta potential was decreased. In 0.1–1 M NaCl and 0.01–1 M CaCl2 solutions, the particle size distribution was wide from a nanometre scale to thousands of nanometres. Zeta potential values were around zero indicating particle aggregation. There was no big difference in silica colloid concentration in sodium chloride or calcium chloride solution. The concentrations of colloids were generally low, however, the more saline the solutions were, and the lower concentrations were due to the aggregation of released colloids. The impact of different ionic media on the stability of colloids released from silica gel is discussed in the context of Olkiluoto conditions.1. P. Hölttä, M. Hakanen, M. Lahtinen, A. Leskinen, J. Lehto and P. Juhola, in Scientific Basis for Nuclear Waste Management XXXII (Mater. Res. Soc. Symp. Proc. Volume 1124, Warrendale, PA, 2009) 1124-Q10-14.2. P. Hölttä, M. Lahtinen , M. Hakanen, J. Lehto and P. Juhola, in Scientific Basis for Nuclear Waste Management XXXIII (Mater. Res. Soc. Symp. Proc. Volume 0, Warrendale, PA, 2009).
9:00 PM - AA6.11
Effects of Solution Chemistry on the Alteration Kinetics of a Simplified Nuclear Glass in Aqueous Medium.
Gabriela Manolescu 1 2 , Odile Majerus 1 , Daniel Caurant 1 , Philippe Barboux 1 , Thibault Charpentier 4 , Francois Devreux 3
1 Laboratoire de Chimie de la Matière Condensée de Paris, CNRS UMR 7574, Chimie-ParisTech (ENSCP), CNRS, Paris France, 2 Agence Nationale pour la Gestion des Déchets Radioactifs, ANDRA, Châtenay-Malabry France, 4 IRAMIS, Service Interdisciplinaire sur les Systèmes Moléculaires et Matériaux, CEA Saclay, Gif-sur-Yvette France, 3 Laboratoire de Physique de la Matière Condensée, Ecole Polytechnique CNRS UMR 7643, CNRS, Palaiseau France
Show AbstractFrom the perspective of a geological disposal of vitrified nuclear waste packages in Callovian-Oxfordian clay, understanding the alteration mechanism of nuclear borosilicate glasses is of considerable importance today, in order to predict their long-term behavior. If a large number of studies has been devoted to the alteration kinetics of glasses in pure water, only few studies exist on the effects of solution chemistry on the leaching kinetics. This study is a contribution to assess the effect of Ca2+, Mg2+, Al3+ ions present in the aqueous solutions, on the alteration kinetics and their mechanisms, in particular on the protective gel layer which forms in the static conditions. These ions are presents in great concentrations in synthetic Callovian-Oxfordian deep groundwater. The alteration kinetics of simplified rare-earth (RE = La, Eu) bearing borosilicate simplified nuclear glasses at 80 °C in different aqueous solutions buffered at pH =7.5 were investigated by static experiments at S/V ratios (glass powder surface area S to leaching solution volume V) of 1 cm-1. It appeared that the presence of calcium in aqueous solutions appreciably modified the leaching kinetics: an increase in the dissolution rate of the glass is observed. However, the alteration of glass in magnesium containing solution remained similar to the one in the pure water. Structural investigations used Raman, IR-ATR and MAS NMR spectroscopies of the altered glasses allow us to discuss the effect of Ca2+, Mg2+, Al3+ ions on the protective gel formed by the recondensation of hydrolyzed species. The evolution of the environment of the rare-earths (RE = Eu) between the glass and the gel is also studied by fluorescence spectroscopy of Eu3+ ions.
9:00 PM - AA6.12
Effect of Molybdenum and Ruthenium on the Crystallization Tendency of a New Nuclear Glass Containing High Rare-earth Concentration.
Nolwenn Chouard 1 2 , Daniel Caurant 1 , Odile Majerus 1 , Jean-Luc Dussossoy 2 , Aurelien Ledieu 2 , Sergei Klimin 3
1 Laboratoire de Chimie de la Matière Condensée de Paris (UMR CNRS 7574), ENSCP Chimie-Paritech, CNRS, Paris France, 2 Laboratoire d’Etude et Développement de Matrices de Conditionnement DEN, MAR/DTCD/SECM, CEA Marcoule, Bagnols-sur-Cèze France, 3 Institute of Spectroscopy, Russian Academy of Sciences, Troitsk (region of Moscow) Russian Federation
Show AbstractVitrification of high level liquid nuclear waste is the internationally recognized method to lower impact on the environment (waste disposal and volume minimization waste). In France, a new confinement glass, aimed not only at immobilizing more concentrated nuclear waste solution than today, arising from the reprocessing of high burn-up-UOX-type nuclear spent fuel, but also at decreasing the number of glass canisters, is currently under study. In this context, high concentration of fission products such as rare-earths, molybdenum and platinoid elements (Ru, Rh, Pd) will be incorporated in this High Level Waste glass (HLW glass) and may induce deleterious effects on its long term behavior (thermal stability, water resistance…). As a matter of fact, one of the major challenges in the optimization of the composition of this new glassy waste form is to avoid crystallization after melt casting in canisters.The aim of this work is to understand crystallization mechanisms by studying, for a simplified aluminoborosilicate glass belonging to the SiO2-Na2O-CaO-Al2O3-B2O3 system, the impact of Nd2O3, MoO3 and RuO2 addition on the competition between the crystallization of apatite Ca2Nd8(SiO4)6O2 and powellite CaMoO4 phases which both may appear in HLW glass during cooling. In this paper, we present the main results on the crystallization tendency of this glass obtained by powder X-ray diffraction (at room temperature and at high temperature) and scanning electron microscopy, after two kinds of thermal treatments: a controlled cooling from the melt (1°C/min), which is representative of the melt cooling rate in industrial nuclear glass canisters, and a thermal treatment of nucleation (2h at Tg+20°C) and growth (30h at 750°C), which is expected to increase crystallization and may thus facilitate characterizations. Moreover, the distribution of Nd3+ cations between the crystalline phases and the residual glass was followed by optical absorption spectroscopy at low temperature. We showed that only heterogeneous nucleation occurs in our glass and that RuO2 clearly acts as a nucleating agent for apatite. Moreover, different crystallization mechanisms occur depending on the thermal treatment of the samples (i.e. controlled cooling from the melt or nucleation and growth from the glassy state). Neodymium and molybdenum cations seem to be very close in the glassy network as Nd2O3 addition stops the phase separation of a molybdate phase in the slowly cooled glass and on the contrary, MoO3 seems to facilitate the crystallization of apatite in the thermal treated glass.
9:00 PM - AA6.13
Molecular Dynamics Simulations of Radiation Damage Cascades in Mixed Alkali Silicate Glasses.
Thorsten Stechert 1 , Michael Rushton 1 , Robin Grimes 1
1 Materials, Imperial College London, London United Kingdom
Show AbstractMixed alkali silicate glasses are used as host materials for long-term immobilisation of high level nuclear waste. The distribution and migration of network modifying species within these wasteforms has significant implications for their long term performance under repository conditions. Here molecular dynamics simulations are used to predict the structural changes that occur in mixed alkali silicate glasses as a result of irradiation. A melt-quench procedure was used to generate glass structures with a composition related to those used for nuclear waste glasses. The effects of irradiation were modelled using the primary knock-on atom (PKA) technique, where large kinetic energies are assigned to single atoms. This allows predictions of radiation damage effects on the distribution and consequently the migration of alkali species in the glass.
9:00 PM - AA6.14
Phase Composition and Elemental Distribution in the Vitrified U-bearing HLW Surrogate.
Sergey Stefanovsky 1 , Boris Nikonov 2 , Boris Omelyanenko 2 , James Marra 3
1 , SIA Radon, Moscow Russian Federation, 2 , IGEM RAS, Moscow Russian Federation, 3 , Savannah River National Laboratory, Aiken, South Carolina, United States
Show AbstractIn the framework of collaboration between SRNL and Daymos/SIA Radon phase composition and elemental distribution in borosilicate glassy material simulating vitrified Sludge Batch 4 (SB4) high level waste (HLW) surrogate were studied. The glass at 55 wt.% waste loading was produced in the demountable cold crucible and cooled to room temperature in the cold crucible using SB4 waste surrogate and commercially available frit 503-R4. Glass samples from a previous pilot-scale cold crucible induction melter (CCIM) campaign were thermally treated in a resistive furnace to simulate the canister centerline cooling (CCC) regime. The blocks were sectioned to investigate phase composition and elemental distribution between various parts of the blocks. Glass blocks were composed of vitreous and spinel structure phases. Spinel was present as both skeleton(dendrite)-type aggregates of fine (micron- or submicron-sized) crystals segregated at early stages of melt solidification and larger (up to tens of microns) individual more regular crystals formed during slow melt cooling. There was some tendency for elemental separation in the glass block produced in the cold crucible with enrichment of the deeper zones with heavier transition metal ions and depletion of Na, Cs, Ca, Al and Si. Uranium was quite uniformly distributed within zone of the block and entered the vitreous phase. IR spectra of the samples from various parts of the block cooled by the CCC regime are nearly same and look like the spectrum of the material from the upper part of the block from the cold crucible. The anionic motif is buil from meta- and pyrosilicate chains and units with trigonally coordinated boron.
9:00 PM - AA6.15
Evaluating Long Term Transport and Accretion of Radionuclide Bearing Dust by Aeolian Processes, Peña Blanca, Chihuahua, Mexico.
Robert Velarde 1 , P. Goodell 1 , M. Ren 1 , T. Gill 1 2
1 Geological Sciences, The University of Texas at El Paso, El Paso, Texas, United States, 2 Environmental Science adn Engineering, The University of Texas at El Paso, El Paso, Texas, United States
Show AbstractThis investigation evaluates potential transport and accretion of dust bearing radionuclides during wind erosion of high-grade uranium ore storage piles at Peña Blanca (50km north of Chihuahua City), Chihuahua, Mexico. The presence of uranium and daughter isotopes in the chain of natural radioactivity will be established. How these isotopes integrate with mineralogy will be investigated. Three sediment collecting stations were deployed: S-1 upwind, S-2 on the ore piles, and S-3 downwind. These dust traps have collected dust particulate since December, 2006 and were disassembled in July, 2009. Moreover, dust swab samples were collected from structures abandoned in 1983 downwind of the stockpiles. Isotope concentrations will be established via gamma-spectroscopy and elemental analysis. At Station S-1 (72 meters west and upwind of the ore piles), the predominant elements detected via electron microprobe analysis are Si, Al, K, Fe, and Ca. Minerals such as quartz, calcite, orthoclase, albite, hematite, and kaolinite were detected via X-Ray Diffraction (XRD). For Station S-2 (on the ore piles), the predominant elements detected via electron microprobe analysis are V, K, U, Si, Al, and Ca. Minerals such as quartz, calcite, orthoclase, albite, kaolinite, and uranophane were detected via X-Ray Diffraction (XRD). For Station S-3 (90.5 meters east and downwind of the ore piles), the predominant elements detected via electron microprobe analysis are Zn, Ca, Si, and Pb. Minerals such as quartz, calcite, orthoclase, albite, hematite, kaolinite, and smithsonite were detected via X-Ray Diffraction (XRD). This study site can serve as an analog to similar uranium mining operations worldwide. These studies have important implications regarding prognosticated uranium mine construction, national security, and public health.
9:00 PM - AA6.2
Kinetic Study of the Reaction Between H2O2 and H2(g).
Joan de Pablo 1 2 , Javier Gimenez 1 , Rosa Sureda 1 , Ignasi Casas 1
1 , Universitat Politecnica de Catalunya, Barcelona Spain, 2 , CTM Centre Tecnologic, Manresa Spain
Show AbstractThe Spent Nuclear Fuel (SNF) dissolution rate decreases in the presence of hydrogen. This decrease could be attributed to the consumption of the oxidizing species formed by the radiolysis of water, which could be catalyzed by the surface of the SNF. One of the most important molecular oxidant identified in spent fuel leaching experiments as a product of the radiolysis of water is hydrogen peroxide. The kinetics of the chemical reaction between hydrogen peroxide and hydrogen has been studied in this work.A hydrogen peroxide solution (concentration around 5xE-5 mol/L) was introduced in an autoclave and put in contact with hydrogen at a known partial pressure. Samples at different intervals of time were taken off and the hydrogen peroxide content was determined. A total of six experiments were carried out, which differed in the hydrogen partial pressure used: 1, 4, 6, 14, 40, and 48 bar.The hydrogen peroxide concentration in solution decreased with time at all the hydrogen partial pressures studied. A hydrogen peroxide consumption rate was calculated considering the decrease of the hydrogen peroxide concentration and time. Between 1 and 14 bar hydrogen partial pressure there was a relatively high increase of the consumption rate with pressure (2E-6 M/d at 1 bar, and 7E-6 M/d at 14 bar), and at pressures higher than 14, consumption rate is shown to be more or less independent on hydrogen partial pressure (around 7,5E-6 M/d).For each hydrogen partial pressure, the decrease of the hydrogen peroxide concentration in solution with time has been modeled considering a second-order reaction kinetics, the fitting of the model to the data has been fairly good. The kinetic constant obtained for the reaction:H2 + H2O2 = 2 H2OHas been determined to be: 3.47E+4 L/mol.s.
9:00 PM - AA6.3
Seismic Tomography Investigation in 140m Gallery in the Horonobe URL Project.
Yutaka Sugita 1 , Hiroyuki Sanada 1 , Takahiro Nakamura 1 , Takao Aizawa 2 , Shunichiro Ito 2
1 , JAEA, Horonobe, Hokkaido Japan, 2 , Suncoh Consultants Co., Ltd., Tokyo Japan
Show AbstractThe Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formations at Horonobe, Japan. The creation of an excavation disturbed zone (EDZ) is expected around the gallery when the gallery is excavated in the underground to dispose the radioactive waste. In-situ excavation disturbance experiment has been performed to determine the rock properties and width of the EDZ in 140m gallery at a depth of 140m below the surface at Horonobe URL. In the experiment, seismic tomography measurement was performed by using seismic source to observe width of the EDZ. Observation area is 3m square horizontal plane along the sidewall of the 140m gallery. During excavation of the 140m gallery, seismic tomography measurement was performed repeatedly with processing of excavation of the gallery, and the change of velocity distribution of the rock around the gallery was observed. It is considered that seismic tomography investigation could catch the created EDZ around the excavated gallery.
9:00 PM - AA6.4
Geo-descriptive Modeling of Water Conducting Features Characterized in Sedimentary Formations in Horonobe Area of Japan.
Koichiro Hatanaka 1 , Doo-Hyun Lim 2 , Eiichi Ishii 1
1 Horonobe Underground Research Center, Japan Atomic Energy Agency, Horonobe-Cho, Hokkaido, Japan, 2 , Golder Associates, Redmond, Washington, United States
Show AbstractA three-dimensional discrete fracture network (DFN) geo-descriptive model is developed for water conducting features (WCFs) in the sedimentary formation of Horonobe underground research laboratory (URL) in Japan. Fracturing and faulting system of the Horonobe URL area is characterized by taking into account i) local geophysical borehole data, ii) regional geologic/structural data, and iii) fracture/fault data (orientation, size) obtained from the surface-based investigations carried out in/around the Horonobe URL area. Volumetric fracture intensity potential is estimated by the correlation and the multi-linear regression analysis of observed data. A regional scale 3-D geo-descriptive DFN model is constructed based on the analyzed fracturing system identified for the water conducting features. The current 3-D geo-descriptive model could be utilized explicitly to derive performance assessment parameters for the hypothetical repository of the high-level radioactive wastes in Japan, and to assist optimization of the safe repository design.
9:00 PM - AA6.5
Immobilization and Long-term Evolution of Selenate in Portland Cement.
Joan De Pablo 1 2 , Isabel Rojo 2 , Miquel Rovira 1 2 , Mireia Grive 3 , Olga Riba 3 , David Garcia 3
1 Chemical Engineering, UPC, Barcelona Spain, 2 Environemental Technology Area, CTM, Manresa Spain, 3 , AMPHOS 21, Barcelona Spain
Show AbstractCements play an important part in the repository designs for the safe disposal of several types of radioactive wastes as they act as a chemical barrier for the retention of radionuclides. Selenium oxyanions are of particular interest because, in nuclear waste management, selenium is considered to have a high priority in the safety assessment. Crystalline calcium sulfoaluminate hydrates (ettringite and monosulfate) seem to be relevant for anion immobilisation by sulphate substitution. In fact, selenate uptake on cement appeared to be related to ettringite through solid-solution formation as has been suggested by Ochs et al. However, the evolution of the hydrated cement phases and their role on the immobilization of selenium at long-term contamination exposure is still under discussion.In the current study, a long-term replenishment batch experiments of 30 cycles (1 cycle: 18 days) have been performed with the aim to simulate a continuous inflow of selenium through a Portland cement phase (CEM I SR). The solution chemical composition after each reaction cycle has been characterized by ICP-OES, ICP-MS and IC. The experimental data indicate a correlation between the sulphate and selenate measured concentrations. Both species are significantly immobilized in the solid phase during the first cycles and, as the experiment progresses, the retention capacity of the cement phase exponentially decreases. The characterization of the cement solid phase by SEM, EDX, XRD and HATR indicates precipitation of secondary ettringite and selenate retained in the newly formed phase. Geochemical modeling of the experimental data is able to explain the correlation between the measured sulphate and selenate concentrations in the aqueous solution and their retention in the solid phase. In this work, we suggest a coprecipitation mechanism of selenate in secondary ettringite as one of the main retention mechanisms of selenium in the studied cement phase.
9:00 PM - AA6.6
Encapsulation of Caesium Loaded Ionsiv in OPC Cement Blends.
Andreas Jenni 1 , Neil Hyatt 1
1 Engineering Materials, University of Sheffield, Sheffield United Kingdom
Show AbstractIonsiv IE-911 (UOP LLC, Des Plaines, Illinois, USA) is used to adsorb cesium radioisotopes from aqueous radioactive solutions. The functional component of this material is a protonated crystalline silicotitanite (CST), which is highly selective for Cs. In the UK, Cs-exchanged CST is considered as an intermediate-level waste but has no scientifically underpinned sentencing and disposal route. Therefore, we have investigated the use of Portland cement based matrices for this purpose with particular reference to potentially deleterious cement / waste interactions, including: decomposition of CST in the high pH cementitious pore solution and release of Cs; release of Cs to the cement pore solution through ion exchange; and desorption of Cs from Ionsiv due to elevated temperatures, which occur during the exothermic cement hydration. In this study, ordinary Portland cement blended with blast furnace slag or fly ash was used to encapsulate Cs-Ionsiv. No morphological indications of Ionsiv grain dissolution was observed by SEM, and the XRD pattern of CST was still clearly distinguishable from the peaks of the cement blends in 28 day samples. In addition, no Cs could be found in any of the cement hydrates. This indicates that CST effectively survives the high pH environment. However, encapsulated CST was found to adsorbed additional ions from the cementitious pore solution including Ca, Na, and K. First results show that these cement ions did not exchange with the Cs adsorbed in CST, because the amount of Cs did not decrease in the same extent, suggesting exchange with the initial protons. Monolithic, static leach tests at 40°C of Cs-loaded Ionsiv encapsulated in cement showed a dependency on the cement system: the fly ash containing sample released approximately twice the amount of Cs compared to the blast furnace slag systems. Nevertheless, 28 day leach fractions measured are 3-10 times lower than comparable measurements carried out on natural zeolites encapsulated in cementitious systems described in literature. The suitability of OPC cement composites for encapsulation of Cs-Ionsiv is discussed.
9:00 PM - AA6.7
Migration Behavior of Alkali Earth Ions in Compacted Bentonite With Iron Corrosion Product Using Electrochemical Method.
Kazuya Idemitsu 1 , Daisuke Akiyama 1 , Akira Eto 1 , Yoshihiko Matsuki 1 , Yaohiro Inagaki 1 , Tatsumi Arima 1
1 Applied Quantum Physics and Nuclear Engineering, Kyushu university, Fukuoka Japan
Show Abstract Carbon steel overpack will corrode by consuming oxygen introduced by repository construction after closure of repository and then will keep the reducing environment in the vicinity of repository. The iron corrosion products can migrate in bentonite as ferrous ion through the interlayer of montmorillonite replacing exchangeable sodium ions in the interlayer. This replacement of sodium with ferrous ion may affect the migration behavior in the altered bentonite not only for redox-sensitive elements but also the other ions. Therefore the authors have carried out electrochemical method, which have been conducted of calcium, strontium or barium with source of iron ion supplied by anode corrosion of iron coupon in compacted bentonite. Fifteen micro liter of tracer solution containing 8.6 M of CaCl2 or 3.0 M of SrCl2 or 1.5 M BaCl2 was spiked on the interface between an iron coupon and bentonite, which dry density was in the range of 1.4 to 1.5 Mg/m3, before assembling. The iron coupon was connected as the working electrode to the potentiostat and was held at a constant supplied potential between - 500 to 300 mV vs. Ag/AgCl reference electrode for up to 2 days. Calcium and strontium could migrate faster and deeper in bentonite specimen than iron in each condition. On the other hand barium could migrate slower than iron. A model using dispersion and electromigration could explain the measured profiles in the bentonite specimens. The fitted value of electromigration velocity was a function of applied electrical potential and 10 to 23 nm/s for calcium, 11 to 19 for strontium, around 5 nm/s for barium and 5 to 10 nm/s for iron, respectively. On the other hand the fitted value of the dispersion coefficient was not a function of applied potential but dry density, and the values were 3 to 8 x 10-12 m2/s for calcium, 2 to 4 x 10-12 m2/s for strontium, about 2 x 10-12m2/s for barium and 2 to 4 x 10-12 m2/s for iron, respectively.
9:00 PM - AA6.8
Influence of Operational Conditions on Retardation Parameters Measured by Diffusion Experiment in Compacted Bentonite.
Ishii Yasuo 1 , Seida Yoshimi 1 , Tachi Yukio 1 , Yoshikawa Hideki 1
1 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency, Muramatsu Tokai, Ibaraki, Japan
Show Abstract
Sorption and diffusion of radionuclides in deep geological environment are the key processes which have been considered in the safe geological disposal of high level radioactive waste. To set reliable parameters for the safety assessment, it is important to establish a method for reliable data acquisition. In a diffusive transportation data acquisition of engineered barrier system, the Reservoir Depletion (RD) test method which is a general data acquisition method for diffusion data acquisition has some potential to cause negative boundary influences, such as mass transfer resistance and change in boundary concentration, to the retardation data in some certain conditions. Thus, it is necessary to understand and to reduce these influences, or to reflect for experimental technique standardization. In the present study, influence of stagnation of test solution and diffusion resistance in the filters on data acquisition for Cs+ and I- were measured by the RD experiment and the range of these uncertainties was estimated by model analysis under the simulated experimental condition.
Kunipia-P was used as a bentonite sample in this study and compacted to a dry density of 1.0 Mg m-3 (20 mm i. d. x 5 mm L). The source reservoir (500 mL) was filled with a NaCl test solution containing 10-7 M of Cs+ and 10-5 M of I- ions. The diffusion period was 30 days for each experiment. The RD curves of Cs+ and I- were determined by ICP-MS measurement. The In-Diffusion (ID) profiles were also obtained by slicing the bentonite sample into 0.5 mm thick pieces and analyzing by ICP-MS after the diffusion period. We investigated operational conditions in this study as to follows; 1) flow rate of the test solution, 2) stirring of the source reservoir, 3) pore size of support porous filter, 4) membrane filter material. The RD experiment showed that the operational conditions affected the RD curves of Cs+ and I- ions. The influence of the operational conditions to the RD curves and ID profiles were also examined theoretically based on a mathematical model simulating the flow of test solution, mass transfer resistance at filter-membrane and membrane-compacted clay interfaces, and sorption on the filter and the membrane, in detail. From the model analysis, it was found that the simultaneous data acquisition and evaluation of RD curves and ID profiles increased the reliability of retardation parameters significantly.
This study was partly financed by the Ministry of Economy, Trade and Industrial of Japan.
9:00 PM - AA6.9
Migration Behavior of Multivalent Radionuclides From Fully Radioactive Waste Glass in Compacted Sodium Bentonite.
Kenso Fujiwara 1 2 , Kazuki Iijima 1 2 , Seiichiro Mitsui 1 , Makoto Odakura 2 , Yukitoshi Kohara 3 , Hiroshi Kikuchi 3
1 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency, Ibaraki Japan, 2 Waste Management Department, Japan Atomic Energy Agency, Ibaraki Japan, 3 , Inspection and Development Company, Ibaraki Japan
Show AbstractIn a repository of high-level radioactive waste, radionuclides will leach from the waste glass and migrate into the surrounding bentonite after very long time. These processes occur simultaneously in the bentonite and should be evaluated to confirm the reliability of individual models and data for the performance assessment of high-level radioactive waste repository.Previous study [1] reported the results of the in-diffusion experiments of Cs in compacted sodium bentonite (Kunigel V1®) in contact with fully radioactive waste glass for 15 to 300 days under aerobic conditions. And Cs migration was successfully interpreted using fundamental one dimensional diffusion model. However, migration of other radionuclides in fully radioactive waste glass were extremely slow because of low solubility, low effective diffusivity and high distribution ratio, especially multivalent elements of actinide and lanthanide.In this study, the similar in-diffusion experiment reported by Ashida et al. [1] was carried out for about 15 years to evaluate the migration behavior of multivalent actinide and lanthanide elements. The bentonite was compacted into a stainless steel cell with 20 mm in diameter and 18 mm in length to produce dry densities of 0.5 and 1.0 Mg m-3saturated distilled water. The form of fully radioactive waste glass was borosilicate glass by using vitrified in CPF (Chemical Processing Facility). The glass sample was sliced into the disc with 4 mm in thickness and sandwiched by two pieces of the saturated bentonite sample in the diffusion cell under aerobic conditions. After 15 years, bentonite sample was sliced and immersed into the HNO3 to extract the radionuclides from the bentonite. Then profiles of Cs, Eu, Pu, Am and Cm in the bentonite sample were evaluated. The concentration profile of Cs in the bentonite was constant due to its high diffusivity.The experimental concentrations of Am, Cm and Eu in contact with compacted sodium bentonite were good agreement with the solubilities calculated by thermodynamic data. On the other hand, the profiles of Am and Cm show two parts with different slopes which cannot be fitted by simple one-dimensional diffusion model considering single specie. Leaching and migration behavior of radio nuclides will be discussed based on the one-dimensional diffusion model considering other mechanism of several species.[1] T. Ashida, et al. Migration behavior of cesium released from fully radioactive waste glass in compacted sodium bentonite. PNC Technical Report, TN8410 98-014(1998).
Symposium Organizers
Katherine L. Smith Embassy of Australia
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Scott Kroeker University of Manitoba
Blas Uberuaga Los Alamos National Laboratory
AA7: Ceramic Wasteforms
Session Chairs
Thursday AM, April 08, 2010
Room 3010 (Moscone West)
9:30 AM - **AA7.1
Synchrotron Radiation XAFS Investigations of Radionuclide Chemistry and Materials Science in Nuclear Waste Forms.
David Shuh 1 , Wayne Lukens 1 , Ponnusamy Nachimuthu 2 , Jonathan Icenhower 2 , Pete McGrail 2 , Suntharampillai Thevuthasan 2 , William Weber 2 , Dennis Lindle 3 , David McKeown 4 , Ian Pegg 4 , Isabelle Muller 4 , Andrew Buechele 4 , Don Paul 5
1 Actinide Chemistry Group, Chemical Sciences Division, The Glenn T. Seaborg Center, Lawrence Berkeley National Laboratory, Berkeley, California, United States, 2 , Pacific Northwest National Laboratory, Richland, Washington, United States, 3 Department of Chemistry, University of Nevada Las Vegas, Las Vegas, Nevada, United States, 4 Vitreous State Laboratory, Department of Physics, Catholic University of America, Washington, District of Columbia, United States, 5 Department of Physics, University of Warwick, Coventry United Kingdom
Show AbstractSynchrotron radiation (SR) methods have been employed to examine and characterize a wide range in nuclear waste and nuclear waste form materials. Over the years, the venerable technique of x-ray absorption fine structure (XAFS) spectroscopy in the hard x-ray region has proven invaluable as a tool to elucidate how actinide elements are incorporated into waste form matrices and to establish the resulting properties of the overall waste form. In particular, XAFS has been useful to determine oxidation states and shown that the utility of analogs for actinides in process development, such as Ce for Pu, is limited. A valuable attribute of XAFS is its complementary nature when coupled to laboratory- and field-based studies with radionuclides For instance, the fundamental and applied chemistry of Tc, a troublesome radionuclide in nuclear waste that poses challenges for long term immobilization, has been largely understood based on a comprehensive set of investigations that relied on XAFS for much of the spectroscopic characterization. Current aspects of understanding Tc incorporation into glass, grout, and possible alternative matrices for Tc will be discussed. Investigations utilizing XAFS in the hard SR spectrum have made large contributions to the understanding of waste form materials chemistry and science, however there have been far fewer soft x-ray SR studies of glasses and pyrochlore materials. Soft x-ray experiments have addressed the speciation of light element constituents of waste form materials such as B, O, Na, Al, Ti, and Fe. In special cases in which a light element like Ti is particularly sensitive to symmetry, valuable structural data may be obtained, and the light element can also serve as an effective monitor of radiation damage. Representative soft x-ray results and their impact will be presented from glass and pyrochlore systems. Future prospects and potential for the application of existing and new SR approaches to probe waste form materials will be highlighted.
10:00 AM - **AA7.2
Defect Properties, Phase Stability and Electronic Properties of Pyrochlores.
Fei Gao 1 , Haiyan Xiao 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractPyrochlore compounds are important host phases for actinide elements in ceramic nuclear waste forms. Once loaded with actinides as a waste form, some pyrochlores will suffer radiation damage from the alpha-decay of actinides and become structurally unstable, that may affect the long term chemical stability for the waste form performance. Recent progress in theoretical approaches to calculate structural and bonding properties, electronic properties and defect formation in a number of pyrochlores is reviewed, which will provide significant insights into the radiation resistance behavior of these materials. Ab initio calculations of A2Ti2O7 (A=Lu, Er, Y, Gd, Sm, Nd, La, Dy, Ho) and B2Sn2O7 (B=Pr, Nd, Sm, Gd, Tb, Ho, Er, Lu, Y) demonstrate that A- or B-site cation ionic radius, position of O48f, lattice constant and the covalency of the bond have a significant effect on the defect formation energies. The results indicate that cation disorder causes local oxygen disordering and that Lu2Sn2O7 is the most resistant to ion-induced amorphization. However, the defect formation energies are not simple functions of the A-site cation radii in A2Ti2O7, in contrast to previous theoretical calculations, and the variation of defect formation energies is in excellent agreement with the radiation-resistance trend of the titanate pyrochlores. Moreover, the effects of f electrons on radiation resistance of A2Ti2O7 will be discussed, which provides an electronic explanation on why Y2Ti2O7 is more radiation resistant than the Dy-and Ho-titanate pyrochlores.In addition, the phase stabilities of Y2Ti2O7 and Y2Zr2O7 under high pressure were investigated by ab initio methods. Y2Ti2O7 has the stable pyrochlore structure, while Y2Zr2O7 has the defect-fluorite structure under ambient conditions. Both the defect-fluorite and defect-cotunnite structures are energetically more stable at high pressure (>40 GPa) in Y2Ti2O7. In the case of Y2Zr2O7, the defect-fluorite phase is not stable under high pressure and it undergoes a structural transformation to the defect-cotunnite state at ~18 GPa. These results are consistent with available experiments.
10:30 AM - **AA7.3
Electrostatic Effects in Nanostructured UO2: Defect Behavior and Radiation Damage.
Blas Uberuaga 1 , Pankaj Nerikar 1 , Simon Phillpot 2 , Susan Sinnott 2 , David Andersson 1 , Xian-Ming Bai 1 , Christopher Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , University of Florida, Gainesville, Florida, United States
Show AbstractUranium dioxide (UO2) is the standard nuclear fuel in pressurized water reactors. Upon burnup within a nuclear reactor, the microstructure undergoes significant evolution, forming very complex structures that depend on the geometry of the fuel. This microstructure, in turn, affects many subsequent properties of the fuel, including fission gas evolution and response to radiation damage events. In this work, we study representative grain boundaries in UO2 in order to understand the qualitative effects they may have on the properties of the fuel. We have found that even the simplest of grain boundaries exhibit complex electrostatic effects, effects that arise from a reconstruction of the atomic structure at the grain boundary, leading to an asymmetry. Depending on the electrical boundary conditions imposed on the material, electric fields may be created within the grain interiors. These electrostatic effects interact strongly with defects to modify their behavior within the material. In particular, they have particular consequence on the formation and evolution of defects produced via irradiation. We discuss the consequences of these results on fundamental mechanisms and engineering implications for future generation nuclear fuels.
11:00 AM - AA7.4
Hydrothermal Stability of Amorphous and Crystalline Pu-doped Zirconolite Ceramics.
Philipp Poeml 1 2 , Joaquin Cobos-Sabate 1 , Thierry Wiss 1 , Philippe Raison 1 , Xavier Deschanels 3 , Thorsten Geisler 2
1 Institute for Transuranium Elements, European Commission, Joint Research Centre, Eggenstein-Leopoldshafen Germany, 2 Institut für Mineralogie, Westfälische Wilhelms-Universität , Münster Germany, 3 Institut de Chimie Séparative de Marcoule, CEA Valrhô Marcoule, Bagnols-sur-Cèze France
Show AbstractZirconolite (CaZrTi2O7) is a promising candidate as a nuclear waste form for the immobilization of actinide radionuclides. One of the most critical properties of a nuclear waste form is its long-term stability in aqueous solutions. In this study we experimentally altered a 239Pu-doped (crystalline), a 238Pu-doped (X-ray amorphous, integrated dose 7*1018 α-decays/g), and a Ce-doped (crystalline) zirconolite under the same conditions. All three ceramics had the same composition, i.e., Ca0.87M0.13ZrTi1.74Al0.26O7 with M = Pu or Ce. This corresponds to ~10 wt.% Pu. XANES measurements on the 239Pu-doped sample revealed that all Pu was present as Pu4+, while EELS and XPS measurements showed that 80% of the Ce was trivalent. The samples had approximately the same size; the Pu-doped samples were discs cut from a pellet, the Ce-doped zirconolite was a specially prepared cuboid with an edge length of 3 mm. Each sample was treated in a Teflon© vessel with 2 ml of 1 M HCl at 200 °C for 3 days (Pu-doped samples) and 14 days (Ce-doped sample).Optical and SEM investigations of the treated samples showed no sign of alteration of the crystalline 239Pu-doped sample, while the amorphous 238Pu-doped sample showed strong alteration features even under the optical microscope. Parts of the disc were covered by a layer of TiO2 crystals and EDX showed that the uncovered areas corresponded to a ZrTiO4 phase where Ca, Pu, and Al were absent. No surface areas yielding the original composition could be found. The cuboid of the Ce-doped sample showed very weak crystal growth on five of its faces, but, surprisingly, the sixth face was covered by a layer of ZrO2 crystals, similar to the layer on the 238Pu-doped sample.The analyses of the aqueous phase by ICP-OES showed that the element concentrations in solution are comparable for the two crystalline samples (239Pu, Ce). In contrast, significantly higher element concentrations were released from the amorphous sample (238Pu) into solution which was expected when considering the higher degree of alteration observed by SEM.Thus, our data clearly reveal that the amorphous sample was altered to a significantly higher degree than the two crystalline samples, suggesting that self-irradiation of the zirconolite lattice has a dramatic effect on its hydrothermal stability. It should be pointed out that we used a strongly acidic solution (initial pH 0) to see any effect within laboratory time scales. It remains a challenge for the future to investigate the change in the alteration kinetics with increasing self-irradiation damage under more realistic physico-chemical conditions as expected in nuclear waste repositories. Another interesting observation of our study is that the oxidation state of Ce and Pu in the zirconolite lattice is different. The III+ ion of Ce is considerably larger than the IV+ ion; this causes effects in the zirconolite structure that affect its stability in aqueous solutions and that are not yet fully quantified.
11:30 AM - **AA7.5
Using Radioparagenesis to Design Robust Nuclear Waste Forms.
Christopher Stanek 1 , Chao Jiang 1 , Blas Uberuaga 1 , Kurt Sickafus 1 , Meiring Nortier 2 , Laura Wolfsberg 2 , Brian Scott 3 , Nigel Marks 4
1 Material Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Chemistry Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 Materials Physics and Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 4 Nanochemistry Research Institute, Curtin University of Technology, Perth, Western Australia, Australia
Show AbstractAlthough public support for the expansion of nuclear power is growing, significant growth is liable to be hindered or even halted by the seemingly intractable nuclear waste problem. A particularly difficult component of the waste problem is that any solution must be highly predictable at time scales not conducive to direct experimental verification. However, we have recently discovered a phenomenon that may permit improved predictability of long-term waste form performance. Specifically, from first principles theoretical methods, we have found that unconventional compounds and crystal structures may form via the chemical transmutation that occurs during radioactive decay (e.g. rocksalt 137BaCl formation from the β- decay of 137CsCl). We refer to this phenomenon as “radioparagenesis.”In this talk, we explore the concept of applying radioparagenesis to the evaluation of long time scale crystalline nuclear waste form performance as well as to the design of chemically robust waste forms. For waste form evaluation, we specifically discuss: (1) the variation in the properties of radioparagenetic phases that govern waste form performance, as compared to conventional materials, (2) the volumetric change associated with radioparagenetic phase formation (using Cs1-xBaxCl as an example), which may lead to waste form cracking and (3) thermodynamic and mechanical stability of radioparagenetic phases. For waste form identification, we introduce the possibility of “backward design,” where the chemical evolution of the waste form is explicitly considered with the goal of achieving a defined end state. The decay of 90Sr to ZrO2 is provided as a demonstrative example. We use density functional theory to examine the performance and design of waste forms. Finally, we discuss the status of accelerated experiments designed to rapidly verify the formation of radioparagenetic phases.
12:00 PM - AA7.6
Fission-product Engineered Waste Forms: The Effects of Decay on Cs-Pollucite.
Tina Nenoff 1 , Terry Garino 2 , Tae-Jin Park 3 , Navrotsky Alexandra 3
1 Surface and Interface Sciences, Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 Electronic & Nanostructured Materials Department, Sandia National Laboratories, Albuquerque, New Mexico, United States, 3 Peter A. Rock Thermochemistry Laboratory, University of California, Davis, California, United States
Show AbstractCesium-137 is a product of uranium fission that has a half-life of 30.2 years and decays to 137Ba through beta decay. CsTiSi2O6.5 is the Cs-containing phase formed by heat treatment of the highly selective and commercially available Cs getter silicotitanate material (UOP IE-910 and IE-911). As a result, CsTiSi2O6.5 has been highly recommended as a waste form for immobilizing radioactive Cs, as an alternative to incorporation in glass. CsTiSi2O6.5 is a zeolite that is an isomorph of pollucite (CsAlSi2O6). Therefore, it is of interest to determine the stability and structure of barium-substituted CsTiSi2O6.5 using techniques such as calorimetry, nuclear magnetic resonance and neutron scattering so that its suitability as a waste form can be assessed. These techniques require that fully crystalline samples over the composition range of interest (up to 50% of Cs with Ba) be synthesized. We report on the synthesis and characterization of barium-substituted CsTiSi2O6.5 materials of two types, CsxBa1-xTiSi2O(7-x/2) and CsxBa(1-x)/2TiSi2O6.5 (1≥x≥0.6), with the pollucite structure. Precursors were chemically synthesized from cesium and barium salts and titanium and silicon organometallics and then thermally treated to induce crystallization. After initial results failed to produce the desired single phase material, a crystal seeding approach was attempted. Seed crystals of CsTiSi2O6.5 were mixed with the precursor powders prior to thermal treatment to provide nuclei to induce crystallization of the desired phase. With the addition of 10 wt% of crystalline CsTiSi2O6.5 to the Ba-containing precursors, nearly single phase pollucite was obtained after 20 hr at 750C for x≥0.6. Microprobe analysis verified that the barium was incorporated in the pollucite structure. These new materials can be used to study the stability of CsTiSi2O6.5 as a durable ceramic waste form which could accommodate both Cs and its decay product, Ba. Initial results of calorimetric, nmr and neutron scattering studies will also be discussed.Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the US DOE’s NNSA under contract DE-AC04-94AL85000.
12:15 PM - AA7.7
Incorporation of Technetium into Ferrite Spinels.
Wayne Lukens 1
1 Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California, United States
Show AbstractOne of the major challenges to the expansion of nuclear energy is disposing of the trace actinides and fission products. Since most of the research on crystalline waste forms has been focused on actinides, considerably less attention has been paid to the fission products. In this regard, technetium is particularly challenging due in large part to the stability (and environmental mobility) of heptavalent technetium. Most of the work on crystalline waste forms for technetium has focused on titanium oxides due to their great chemical durability and the similar ionic radii of tetravalent titanium and technetium. However, incorporation of tetravalent technetium into titanium oxides can be challenging due to differences in redox behavior as well as the different aqueous chemistry of tetravalent titanium and technetium. The other common element with an ionic radius similar to tetravalent technetium is iron (high-spin, trivalent iron has the same ionic radius as tetravalent technetium). While iron oxides are less durable than titanium oxides, they are still highly durable, especially hematite and trivalent ferrites, and potentially could be waste forms for technetium. As a first step to investigating the utility of iron oxides as technetium waste forms, incorporation of tetravalent technetium into ferrite spinels from aqueous solutions of pertechnetate was examined. The materials were characterized by a combination of x-ray diffraction and x-ray absorption fine structure spectroscopy. While these materials are not waste forms in and of them themselves (they would need to be consolidated into a low surface area morphology), incorporation of technetium into the iron sites of these crystalline oxides is straightforward.
12:30 PM - AA7.8
Computational and Experimental Investigations of the Structure of Halite Nanoparticles.
Ahmed Ismail 1 , Martin Nemer 1 , Dennis Powers 2
1 Performance Assessment and Decisions Analysis, Carlsbad Programs Group, Sandia National Laboratories, Carlsbad, New Mexico, United States, 2 Department of Geology and Geological Engineering, University of Mississippi, University, Mississippi, United States
Show AbstractSalt repositories are currently being utilized for the disposal of nuclear waste. One issue that deserves further study is the effect of radiation damage on halite. Radiolysis of halite can cause changes in the chemical interactions between the host rock and the waste. Observing the impact of electron-beam damage on halite during Transmission Electron Microscopy (TEM) is a simple, in situ method of investigating the effects of heating and radiolysis on halite crystallography, texture, and chemistry. Understanding these effects can lead to improved, integrated process models describing the macroscale thermal and mechanical response of halite to stresses due to radiation.Recently, we have developed techniques for preparing and imaging halite samples using Transmission Electron Microscopy. These techniques preserve orientation, fluid inclusion geometry, and texture. While imaging halite samples in the TEM we have observed large (200 nm) truncated-octahedral halite nano-particles, formed as the result of TEM electron-beam damage. While surface-energy versus lattice-strain competition resulting in non-cubic particles has been observed many times before for metals, these particles are unusual because of their large size. Using molecular dynamics simulations of salt systems comprising thousands of atoms, we are studying the observed crystal structure and morphology as a function of system size and initial conditions. This work will add fundamental knowledge to the behavior of salt under radiation, which may be helpful in the design of future high-level waste repositories. Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE AC04 94AL85000. This research is funded by WIPP programs administered by the Office of Environmental Management of the U.S Department of Energy.
12:45 PM - AA7.9
Cation Ordering in β-Tricalcium Phosphate.
Eleanor Jay 1 , Emily Michie 1 , Shirley Fong 2 , Phillip Mallinson 2 , Lee Gerrard 2 , Robin Grimes 1 , Brian Metcalfe 2
1 Materials Department, Imperial College London, London United Kingdom, 2 Materials Science Department, AWE, Aldermaston United Kingdom
Show Abstractβ-TCP, Ca3(PO4)2, has a primitive unit cell consisting of twenty one formula units and five different cation sites. One of the cation sites (Ca(4)) is reported as half occupied, leading to many possible arrangements of cations over sites. Here we present on the energies and symmetries of different configurations, with the aim of identifying particularly stable structures. In the (1x1x1) single unit cell there are six Ca(4) sites, only three of which are occupied, leading to 20 different configurations. However, in order to investigate possible configurations in the full hexagonal symmetry a (3hx1x1) super cell was investigated; this gives rise to nearly 50,000 different configurations. The lowest energy configurations identified in the (3hx1x1) cell have symmetry groups P31C23 and P31C33, which are isomorphic subsets of the R3c (1x1x1) lowest energy unit cell, which cannot be generated from this unit cell.X-ray diffraction patterns generated at a range of temperatures from pure samples of β-TCP are also compared to simulated models. In all cases the energies of lattice structures were calculated using a classical interatomic pair potential model.
AA8/BB7: Joint Session: Radiation Effects in Wasteforms and Ceramics
Session Chairs
Ram Devanathan
Karl Whittle
Thursday PM, April 08, 2010
Room 3010 (Moscone West)
2:30 PM - **AA8.1/BB7.1
Radiation Effects in Actinide-containing Ceramics for the Advanced Fuel Cycle.
William Weber 1 , Ram Devanathan 1
1 Fundamental & Computational Sciences Directorate, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractAdvanced nuclear fuel cycles may include a separate waste stream for minor actinides that may be immobilized directly in waste forms or in inert matrices for in-reactor transmutation and subsequent geologic disposition. Since some minor actinides have relatively high spontaneous fission rates or neutron fission cross sections, radiation effects from both alpha decay and fission may need to be considered. Self-radiation from alpha decay and fission of actinides in crystalline ceramic host phases generally leads to track formation, point defect accumulation, and crystalline phase transformations or amorphization, which may be accompanied by macroscopic swelling and increases in dissolution rates. The results of computer simulations, new models, and experimental studies in several relevant materials using short-lived actinides and ion-beam irradiation methods, along with comparisons to natural mineral data, will be presented to highlight the fundamental understanding and models of radiation effects in actinide-containing ceramics developed over the past 30 years. Low-energy (~MeV) heavy-ion irradiation techniques have been demonstrated to be very effective in simulating radiation effects from alpha decay over a large range of experimental conditions in order to develop more detailed scientific understanding and predictive models. Swift-heavy ion irradiations (~0.1 to 2 Gev) provide a means to better understand the nature of fission tracks. The integration of computer simulations into these studies have advanced the interpretation of experimental results and led to the development of a comprehensive atomic-level understanding of radiation damage processes and predictive models of the complex evolution of radiation damage in actinide-containing waste forms or inert matrix fuels with time and temperature.
3:00 PM - AA8.2/BB7.2
Quantification of α-type Radiation Damage in Zircon.
Katie Gunderson 1 , Clive Brigden 1 , Eric Vance 2 , John Hanna 2 , Ian Farnan 1
1 Department of Earth Sciences, University of Cambridge, Cambridge United Kingdom, 2 , Australian Nuclear Science and Technology Organisation, Menai, New South Wales, Australia
Show AbstractQuantification of α-radiation damage in materials is important because of the large number of heavy α-emitters in nuclear waste and nuclear materials. Each α-decay event causes radiation damage by two processes: the emission of the α-particle itself, and the recoil of the heavy nucleus. The majority of localized structural damage leading to amorphization is caused by the recoil of the heavy actinide nucleus (70-100 keV), such as that of plutonium, americium or curium. The higher energy α-particle (4.5-5.5 MeV) mainly causes ionizations, but is also known to cause some atomic displacements as it is slowed by collision with atomic nuclei. Nuclear magnetic resonance (NMR) has previously been used to quantify the cumulative amount of radiation damage in material due to the total damage event (heavy recoil + α). This study attempts to quantify the damage due to α-particles alone and provides precise constraints on the number of damage events through 7Li spin counting. Here we present results from zircon synthesized with natural boron as B2O3 inclusions. The natural abundance of 10B is 19.9%, and it undergoes the reaction 10B + n → 7Li + α. Samples were neutron-irradiated at HIFAR reactor at Australian Nuclear Science and Technology Organisation for one day, one week, and one month at a neutron fluence of 1013 cm2/s. The amorphous fraction of the sample is quantified using 29Si NMR by comparing the area of a narrow line corresponding to undamaged zircon with the area of a broad resonance at more negative chemical shifts corresponding to amorphized material. A calibration curve is constructed for 7Li NMR signal as a function of the number of 7Li atoms, and is used to quantify the number of 7Li atoms in each irradiated sample. Complete fission occurred in the sample that had been irradiated for one month, and analysis of the 29Si NMR spectrum indicates that 30% of the structure had been damaged. Each 10B + n → 7Li + α reaction creates two energetic light particles that we consider together. The amount of 7Li detected in the 1 month sample is 4.08 x 1018 atoms, which translates to 1.95 x 1019 α events / g material. In the same sample the 29Si NMR spectrum shows a damage level of 5.90 x 1021 atoms/g. This leads to a damage rate of 302 atoms/α. Subsequent experiments on lower levels of irradiation should provide better constraints on the damage per α. Also, X-ray diffraction will allow changes in unit cell parameters to be quantified as a function of ‘α-particle’ only damage.
3:15 PM - AA8.3/BB7.3
Ion-irradiation-induced Damage Evolution in Titanate Pyrochlores.
Yanwen Zhang 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractPyrochlore materials, due to the remarkable elemental versatility in the A2B2O7 crystal structure, are considered for a wide range of applications, such as fuel cells, catalysts, inert fuel matrices, and waste forms for the immobilization of actinides. Considerable self-radiation damage from alpha-decay in actinide-bearing phases can result in amorphization, macroscopic swelling and significant increases in dissolution rates, and these changes in structure and chemical durability affect long-term performance of the actinide waste forms. Irradiation effects in Ho2Ti2O7 and Sm2Ti2O7 single crystals were studied using 1.0 MeV Au+ ion at room temperature, which provides a reasonable simulation of the damage evolution behavior due to alpha recoils. Damage evolution as a function of irradiation dose was investigated by Rutherford backscattering spectroscopy along the <001> direction. A disorder accumulation model, with contributions from the amorphous fraction and the crystalline disorder, has been fit to the damage accumulation data. The damage evolution behavior indicates that the relative disorder on the Ho sublattice follows a nonlinear dependence on dose and that defect-stimulated amorphization is the primary amorphization mechanism, which is similar to the irradiation behavior previously observed in Sm2Ti2O7. A slower observed damage accumulation rate for Ho2Ti2O7, as compared with damage evolution in Sm2Ti2O7, is mainly attributed to a lower effective cross section for defect-simulated amorphization. The critical dose for amorphization under 1.0 MeV Au+ irradiation in Ho2Ti2O7 and Sm2Ti2O7 crystals are in good agreement with TEM results for polycrystalline Sm2Ti2O7 and Gd2Ti2O7 under 0.6 MeV Bi+ irradiation and for 244Cm-doped Gd2Ti2O7, indicating that amorphization due to heavy-ion energy deposition is relatively independent of dose rate and ion mass near room temperature. During thermal annealing in an 18O environment, the increase in exchange near the critical temperature for amorphization between 16O in the Sm2Ti2O7 sample and 18O suggests a correlation between oxygen vacancy mobility and the critical temperature.
3:30 PM - AA8.4/BB7.4
Experimental Insight into the Radiation Resistance of Zirconia-based Americium Ceramics.
Renaud Belin 1 , Philippe Martin 1 , Philippe Valenza 1 , Andreas Scheinost 2
1 DEN - DEC - SPUA - LMPC, CEA Cadarache, Saint Paul Lez Durance, 14999, France, 2 Institute of Radiochemistry, Forschungszentrum Dresden-Rossendorf (FZD), Dresden Germany
Show AbstractCeramics intended for use as nuclear fuels, transmutation targets or actinide immobilization matrices have to endure severe conditions including internal radiation. Resistance to amorphization and limited swelling are therefore key requirements. While zirconia-based materials with defect-fluorite structure have shown high tolerance against external ion-beam irradiation, few experimental studies have demonstrated that these structures also resist under more realistic conditions, i.e. with homogeneous internal radiation from α-emitting actinides within the structure. Here, we provide for the first time experimental insight into the radiation-resistance mechanisms of americium pyrochlore 241Am2Zr2O7. The presence of 62 wt% of 241Am (t1/2=432 y) in the structure provided a high dose rate which enabled us to follow within only four years an α self-irradiation-induced aging process corresponding to much longer time periods in the case of immobilization ceramics [1]. X-ray diffraction (XRD) and X-ray absorption spectroscopy (XAS) analyses were used to selectively probe the long-range and the short-range order and the local environment of both cations [2,3].The phase transition from the pyrochlore to the defect-fluorite structure was accompanied with an unusual negative lattice expansion. Once the fluorite structure was reached, neither volume changes nor amorphization were observed over a time course of 4 years. The cumulated dose during this period was 9.4×1018 α-decay events.g-1; corresponding to 0.80 dpa. This is equivalent to the dose accumulated by multiphase ceramics with 20wt% 239Pu during 1,000 years of storage [1]. The defect-fluorite structure is therefore remarkably resistant to such radiation doses.Combining XRD Rietveld refinement and XAS analysis allowed a thorough understanding of how lattice point defects are formed and recombined during the order-disorder transition. The disorder relaxation proceeds through the simultaneous formation of cation antisites and oxygen Frenkel pairs, in line with former molecular dynamics studies. Moreover, EXAFS analysis revealed a disruption in the long-range order and a markedly different development in the local environments of zirconium and americium: while Am-O polyhedra show an increasing disorder with increasing exposure, the Zr-O polyhedral units remain highly ordered. However, they rotate along edges and corners, thereby reducing the structural strain imposed by the growing disorder around americium. We believe it is this particular property of the compound that provides the remarkable resistance to radiation, making it attractive as a durable actinide immobilization host and as a stable transmutation target.[1] Weber W. et al. 1998 J. Mater. Res. 13 1434-1484[2] Martin P. et al. 2007 J. Alloy. Compd. 444-445 410-414.[3] Belin R.C. et al. 2009 Inorg. Chem. 48(12) 5376-5381
3:45 PM - AA8.5/BB7.5
Ion Beam Irradiation of Lanthanum Compounds in the Systems La2O3-Al2O3 and La2O3-TiO2.
Karl Whittle 1 , Gregory Lumpkin 1 , Mark Blackford 1 , Katherine Smith 1 , Nestor Zaluzec 2
1 Materials Engineering, ANSTO, Menai, New South Wales, Australia, 2 Electron Microscopy Center, Argonne National Laboratory, Chicago, Illinois, United States
Show AbstractThin crystals of La2O3, LaAlO3, La2/3TiO3, La2TiO5, and La2Ti2O7 have been irradiated in situ using 1 MeV Kr2+ ions in the Intermediate Voltage Electron Microscope-Tandem User Facility (IVEM-Tandem), at the Argonne National Laboratory (ANL). We observed that La2O3 remained crystalline to a fluence greater than 3.1 × 1016 ions cm-2 at a temperature of 50 K. The four binary oxide compounds in the two systems were observed through the crystalline-amorphous transition as a function of ion fluence and temperature. Results from the ion irradiations give critical temperatures for amorphisation (Tc) of 647 K for LaAlO3, 840 K for La2Ti2O7, 865 K for La2/3TiO3, and 1027 K for La2TiO5. The Tc values observed in this study, together with previous data for Al2O3 and TiO2, are discussed with reference to the phase diagrams for the La2O3-Al2O3 and La2O3-TiO2 systems and the different local environments within the four crystal structures. Results suggest an observable inverse correlation between Tc and melting temperature (Tm) in the two systems. More complex relationships exist between Tc and crystal structure, with the stoichiometric perovskite LaAlO3 being the most resistant to amorphization
4:30 PM - **AA8.6/BB7.6
Response of Complex Oxides to Extreme Conditions of Irradiation and Pressure.
Maik Lang 1
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractComplex oxides such as A2B2O7 and A2BO5 exhibit a variety of properties that find application in a number of different technologies, from electrolytes in solid oxide fuel cells to actinide-bearing phases. Ion beam irradiations (energy: GeV) have been used to systematically study the effects of pure electronic excitation and ionization in Gd2Zr2-xTixO7 and Gd2TiO5. Synchrotron X-ray diffraction, Raman spectroscopy, and transmission electron microscopy (TEM) revealed radiation-induced phase transitions that include the crystalline-to-amorphous and the order-disorder structural transformations, as well as the formation of high-temperature and high-pressure phases. The extreme conditions prevailing in a single ion track trigger the formation of equilibrium- and non-equilibrium phases, and entire phase diagrams are accessible at the nanoscale. The specific type of material modification can be manipulated by changing the target composition (e.g., Ti-content), beam parameters (e.g., energy loss dE/dx), and external irradiation condition (e.g., pressure and temperature). As an important application, we show that the combination of relativistic heavy ions and high pressure can result in the formation and stabilization of a new metastable Gd2Zr2O7 phase that cannot be obtained by irradiation or pressure applied separately. TEM and quantum-mechanical calculations suggest that these novel structural modifications are caused by the formation of nanocrystals that change the energetics of the phase transformations. This result highlights the importance of the combined use of high pressure and high-energy ion irradiation as a new means for manipulating and stabilizing novel materials to ambient conditions that otherwise could not be recovered.
5:00 PM - AA8.7/BB7.7
Comprehensive Description of the Ion Irradiation-induced Microstructural Modifications in Nuclear Materials: Fluorite-structured Oxides as a Test Case Study.
Aurelien Debelle 1 , Frederico Garrido 1 , Lionel Thome 1
1 CSNSM - CNRS, Université Paris-Sud 11, Orsay Cedex France
Show AbstractMaterials for nuclear applications are inherently submitted to harsh radiative environments. In particular, they must face irradiation with various types of ions in a broad energy range, from low-energy heavy recoil nuclei arising from the alpha decay of actinides, to swift fission fragments. In the majority of the cases, ion irradiation induces damage in the material, which may be deleterious for its intrinsic physical and chemical properties for which it has been chosen. In the case of low- and medium-energy (tens of keV to few MeV) ion irradiation, it has been shown by RBS/C that the damage buildup occurs in several steps as a function of the accumulated ion fluence [1]. Each step is characterized by the presence of a predominant type of defects. The transition from one step to the following one seems to be triggered by radiation-induced defect re-organization which leads the material to a more stable energetic state [2]. Therefore, the ability to monitor the microstructural transformations and to determine the key parameters that trigger them definitely appears as important issues for a comprehensive knowledge of the behavior of nuclear materials under irradiation. The radiation-induced defects generate elastic strains and stresses which can in turn play a role on the defect behaviour. In the case of crystalline structures, XRD is one of the most relevant techniques to determine the strain/stress state of materials. It is here demonstrated that XRD measurements combined with RBS/C and TEM experiments allow getting a comprehensive description of the behavior of nuclear materials under irradiation. To illustrate this statement, two test case nuclear materials, namely cubic zirconia and urania, have been studied. XRD allowed determining both the strain/stress state of irradiated materials, and the variation of the strain with the irradiated depth. These depth strain profiles have been compared to SRIM-predicted and experimental damage (channeling) and implanted specie (RBS) distributions. Furthermore, the possibility to monitor the damage level as a function of the ion fluence has been established [3]. A multi-step damage buildup, similar to the one obtained by RBS/C, has been found. Defect clusters inducing large strains are formed in the first step, while dislocation loops develop at higher fluences to relieve the stored elastic energy. Thus, this elastic energy term seems to drive microstructural transformations under irradiation. [1] K.E. Sickafus et. al., J. Nucl. Mater. 274, 66 (1999).[2] J. Jagielski et. al., Appl. Phys. A97, 147-155 (2009).[3] S. Moll et. al., J. Appl. Phys. 106, 073509 (2009).
5:15 PM - AA8.8/BB7.8
Real-time Observation of Radiation Damage in Zircon by High-energy Electron Beams.
Nan Jiang 1
1 Physics, Arizona State University, Tempe, Arizona, United States
Show AbstractRadiation effects in zircon have been extensively studied because of the potential application of zircon for immobilization of plutonium. In natural zircon, which contains traces of uranium and thorium, α-decay events degrade the zircon structure resulting in a crystalline-to-amorphous transformation. The same transformation has also been observed and studied in Pu-doped zircon, neutron irradiated and ion-beam irradiated zircon. The amorphization process is generally believed to result from the spontaneous local collapse of the crystal structure due to a local high defect concentration. The direct observations of the crystalline-to-amorphous transformation were based on transmission electron microscopy (TEM) observations. In high-resolution electron microscopy (HREM) images, regions with mottled contrast were observed, and attributed to amorphous or disordered volumes in the lightly damaged zircon, while small crystallites were seen in an amorphous matrix in the heavily damaged zircon. Most interestingly, the HREM images showed that microstructural evolution in the self-irradiated natural zircon was very similar to that in the ion-beam damaged zircon, although the damage accumulation rate in the former was several orders of magnitude lower than that in the ion-beam irradiation experiments. The conclusions largely supported and encouraged the use of ion-beam irradiation to simulate the α-decay and fission track damage effects in zircon. Nevertheless, the atomic-level structural changes and the structure of damaged “amorphous” zircon are still not well understood. In review of various data obtained from a variety of analytical techniques, information about the composition of “amorphous” zircon, at least at the nanometer scale, was missing. Does heavily damaged zircon have the same composition as undamaged zircon? Therefore radiation damage by high-energy (200 keV) electrons in zircon has been studied thoroughly using imaging, diffraction and electron energy-loss spectroscopy (EELS) techniques in TEM. Both structural and compositional changes during the damage were measured using the above techniques in real time. It is found that the electron-beam induced damage in zircon is more or less similar to that induced by ion-beam irradiation or even α-decays in natural or Pu-doped Zircon. The appearance of the patches is very similar to that observed as mottled diffraction contrasts in α-decay-damaged and ion-beam irradiated zircon. However, the damage by electron beam was mainly caused by the preferential sputtering of O. The loss of O occurred initially within small sporadic regions with dimension of several nanometers, resulting in the direct transformation of zircon into ZrxSiy. These isolated patches gradually connect each other, and eventually cover the whole area of the electron beam.
5:30 PM - AA8.9/BB7.9
Key Role of the Cations Interstitial Structure in the Radiation Resistance of Pyrochlores.
Alain Chartier 1 , Gilles Catillon 2 , Jean-Paul Crocombette 3
1 , CEA-Saclay, Gif-Sur-Yvette France, 2 , Université Paris-Est, G2I, Marne la Vallée France, 3 DEN/DANS/DMN/SRMP, CEA-Saclay, Gif-Sur-Yvette France
Show AbstractAmong the structurally related fluorite compounds, the pyrochlore oxides (A2B2O7) have many technological applications due to the variety of their properties. A fundamental understanding of their response to irradiations is also of utmost interest for the nuclear industry as they can be used as an inert matrix. We focus here on the Gd2ZrxTi2-xO7 (0 < x < 2) solid solution as it is a textbook example of the response of pyrochlores under irradiations. Its response varies widely with composition, from no amorphization in Gd2Zr2O7 at any temperature (down to 30 K) to complete amorphization in Gd2Ti2O7 up to a so-called critical temperature, beyond which no amorphization is possible.We show in the present work [1] that continuous Frenkel pairs’ (FP) accumulation within molecular dynamics can quantitatively reproduce the irradiation response of the whole solid solution Gd2ZrxTi2-xO7, as a function of composition and temperature. The calculated critical temperatures TC for amorphization are in quantitative agreement with the experimental measurements.In fact, the decrease of the radiation resistance as a function of the increase of Ti content in Gd2ZrxTi2-xO7 is shown to be driven by a change in the configuration of the B interstitials. The stabilization of Ti-Ti dumbbells in the partially disordered fluorite structure hinders their recombination with vacancies. Conversely, the isolated Zr interstitials readily recombine with vacancies. So, as the amount of Ti increases in the B site of Gd2B2O7, the stability of the B-B dumbbells increases in the fluorite structure and prevents the reconstruction of the structure which eventually collapses to the amorphous state. Present calculations also unify the previous ground states based criteria from Sickafus [2] and introduce the recombination of the B-B dumbbells as the missing phenomenon in the rate equations.References [1] A. Chartier, G. Catillon, and J.-P. Crocombette, Phys. Rev. Lett. 102 (2009) 155503.[2] K. E. Sickafus, L. Minervini, R.W. Grimes, J. A.V. Ishimaru, F. Li, K. J. McLellan, and T. Hartmann, Science 289, 748 (2000).
5:45 PM - AA8.10/BB7.10
In-situ Observations of Ion-induced Phase Transformations Using a Reflection High-energy Electron Diffraction (RHEED) System.
Jonghan Won 1 , James Valdez 1 , Igor Usov 1 , Kurt Sickafus 1
1 Materials Science and Technology Divison, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractA novel method was developed to monitor in-situ ion irradiation damage in crystalline materials. The new method uses reflection high-energy electron diffraction (RHEED) to collect electron diffraction patterns while performing ion irradiation. In our RHEED experiments, we use a ~50 keV electron beam impinging on a crystalline substrate at grazing incidence. The diffracted electrons are collected on a phosphor screen with a cold cathode device. In this presentation, we report on ion irradiations of a single crystal (LaAlO3). Irradiations were performed using 100 keV Ar ions at room temperature. Before the single crystal was bombarded with Ar ions, a diffraction pattern with strong diffraction maxima was observed. During the Ar ion irradiation, electron diffraction patterns were collected from the near surface region (continuously). The initially strong diffraction patterns gradually faded in intensity until no diffraction spots were visible on the phosphor screen at the fluence of 6-8x1014 ions/cm2. To confirm the presence of the ion-induced amorphization in the near surface region, we used grazing incidence X-ray diffraction (GIXRD) and electron diffraction in a transmission electron microscope (TEM). Both characterization techniques corroborated the RHEED in-situ measurement, showing the existence of an amorphous region produced by the 100 keV Ar ions. Results from this investigation showed that by monitoring RHEED patterns during ion irradiation, it is possible to determine precisely threshold conditions for changes in material crystal structure, as a function of ion irradiation fluence, temperature, etc.
Symposium Organizers
Katherine L. Smith Embassy of Australia
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Scott Kroeker University of Manitoba
Blas Uberuaga Los Alamos National Laboratory
AA9: Solubility of Nuclear Material
Session Chairs
Friday AM, April 09, 2010
Room 3010 (Moscone West)
9:30 AM - **AA9.1
Factors Impacting the Corrosion Behaviour of Spent Fuel and Other High Level Waste Forms.
Vincenzo Rondinella 1 , Thierry Wiss 2 , Joaquin Cobos 3 , Detlef Wegen 3 , Paul Carbol 3 , Daniel Serrano-Purroy 3 , Emilio Maugeri 2 , Daqing Cui 4
1 Hot Cells, JRC-ITU, Karlsruhe Germany, 2 Materials Research, JRC-ITU, Karlsruhe Germany, 3 Nuclear Chemistry, JRC-ITU, Karlsruhe Germany, 4 , Studsvik AB, Nykoping Sweden
Show AbstractThe corrosion behaviour of spent fuel and other high level ceramic waste forms placed in a geologic repository for final disposal will be determined by the combined effect of many factors acting and developing over an open-ended timescale.The history of the fuel (initial composition, irradiation parameters, burnup, age) determines fission product and actinides inventory and distribution in the material. These, in turn, determine the specific radioactivity of the fuel due to decay. The decay processes cause damage in the crystalline structure of the fuel, which superimposes to the fission damage that had occurred in pile and evolves with the ageing of the waste material. Alpha-decay, in particular, generates helium, whose build-up after high cumulated dose/long storage times may have significant effects on the spent fuel structure. Both in pile and storage histories (including the chemico-physical conditions of storage) also affect the structure morphology evolution of the fuel. The specific surface area of a fuel with a given history determines the amount of radionuclides exposed to water corrosion in case of the first contact between waste and groundwater in the repository. The chemistry conditions at the surface (oxidation state, phase composition) together with the radiolytic process caused by the specific activity of the fuel affect the amount of material that can be dissolved in groundwater over a given time interval. The fuel properties act in combination with local environment quantities like: composition of groundwater (ionic species, colloids); pH, temperature, redox conditions/coupling (overall and locally); amount, composition and specific surface of solid phases (both natural minerals and engineered containment materials); other important phases, like e.g. the hydrogen phase building up in deep geologic repositories in association with water corrosion of iron canister materials. Mechanisms like e.g. catalytic reactions associated with metallic precipitates and surface precipitate on spent fuel and other waste forms, or coupling between species/phases expected to be present in the repository environment may also affect the corrosion behaviour of the waste material. This presentation shows highlights related to the above-mentioned factors from studies on irradiated fuels and analogue materials (both natural and tailor-made) performed at ITU in collaboration with various partners and in the frame of international projects. The final goal of these studies is to assess the long-term stability of spent fuel and other waste forms in view of the implementation of final disposal concepts in Europe.
10:00 AM - AA9.2
Studies on Commercial Spent Nuclear Fuel Corrosion: Effect of Sample Morphology and Radial Position in IRF Determination.
Joan De Pablo 1 4 , Daniel Serrano-Purroy 2 , Ernesto Gonzalez-Robles 4 2 , Jean-Paul Glatz 2 , Detlef Wegen 2 , Ignasi Casas 1 , Javier Gimenez 1 , Frederic Clarens 4 , Aurora Martinez-Esparza 3
1 Chemical Engineering, UPC, Barcelona Spain, 4 Environmental Technology Area, CTM, Manresa Spain, 2 , JRC-ITU, Eggenstein-Leopoldshafen Germany, 3 , ENRESA, Madrid Spain
Show AbstractThe denominated IRF is considered in Performance Assessment (PA) exercises to govern the dose that could arise from the repository. The IRF concept can be defined as all RN fractions that will be instantly released or more correctly during a short period of time when water reaches the fuel. A possible definition of IRF comprises the total inventory of radionuclides located in the gap between the fuel and cladding, at the grain boundaries and in the high burn-up structure (HBS) when this phase is present. In addition to the difficulties in a practical definition of IRF, the experiments focused on this measurement are scarce, especially for high burn-up UOX fuels.With increasing burn-up (BU), neutron capture of U-238 produces Pu-239 leading to an external layer with a higher BU, increased porosity and fuel grain subdivision, the so-called high burn-up structure (HBS). The depth of this layer grows with the BU and depends on the irradiation history. This layer is observed for BU’s higher than 40 GWd/tU. No experimental data can be found regarding the dissolution rates of this region independently from core data.For these reasons, in the present work static leaching experiments were carried out for two commercial spent UO2 fuels of 48 and 60 GWd/tU. In both studied fuels four different samples were studied: one prepared from the center of a pellet, a second one prepared from the periphery, enriched with HBS, a third one consisting on remaining HBS attached to the cladding and, finally, a complete pellet. All four samples were leached in bicarbonate water as leachant.
10:15 AM - AA9.3
Complex Antimonates for Nuclear Fuel and Waste Management.
Dmitry Zakharyevich 1 , Vladimir Burmistrov 1 , Aleksey Rekunov 1
1 Condensed Matter Physics, Chelyabinsk State University, Chelyabinsk Russian Federation
Show AbstractThe existing nuclear fuel reprocessing facilities use a technique based on PUREX process. Huge volumes of liquid low- and intermediate-level wastes are generated during this process. If a material used for extraction of hazardous isotopes could be almost directly used also for immobilization, it would greatly improve the safety and reduce the cost of the process. Crystalline polyantimonic acid (CPAA) is among the best inorganic ion-exchangers, and has wide application in extraction methods. Its ion-exchange capacity and selectivity could be easily tuned by chemical modification. On the other hand, CPAA and some of its derivatives crystallize into pyrochlore structure that possesses unique radiation-resistance properties. Although pyrochlore CPAA and its salts (antimonates) are not stable, our results had shown that appropriate doping of skeleton of pyrochlore structure (e.g. with oxides of W, Mo, other d-elements) makes it stable under various conditions. The results of the studies of ion-exchanging properties of CPAA and its derivatives in imitating solutions and thermodynamic stability of doped antimonates are presented in this report with a scheme of new reprocessing technique, in which the extraction and immobilization stages merge in one.
10:30 AM - AA9.4
The Fate of Radiogenic Iodine During the Electrochemical Treatment of Spent EBR-II Fuel.
Steven Frank 1
1 Pyroprocessing Technology, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractRadiogenic iodine produced during the fissioning of nuclear fuel is one of the more difficult radionuclides to capture and immobilize during aqueous treatment of spent fuel. However; fission-product iodine is retained in the process stream during the electrochemical and pyrometallurgical treatment of metallic spent nuclear fuel. Electrochemical treatment of Experimental Breeder Reactor-II (EBR-II) fuel at the Idaho National Laboratory first involves chopping fuel elements into small segments, then anodically dissolving the fuel into the electrorefiner molten salt that is composed of LiCl-KCl. Any iodine that escapes the fuel material is believed to become incorporated into the bond sodium that surrounds the fuel in the fuel rod. The majority of the bond sodium is introduced into the electrorefiner with the chopped fuel elements. The chemically active fission products in the fuel and bond sodium are oxidized along with the fuel and dissolve into the electrorefiner electrolyte. Eventually, the electrolyte must be disposed of, and this is accomplished by immobilizing the electrorefiner salt in a zeolite/glass-binder matrix and processing into a ceramic waste form for geological disposal. Discussed in this presentation are the fission-product iodine measurements performed on the bond sodium of the metallic fuel prior to electrorefining and on the electrorefiner salt. Furthermore, antidotal evidence suggests that radiogenic iodine remains in the ceramic waste form (CWF) after processing. Retention of iodine in the CWF has been demonstrated by analyzing non-radioactive, surrogate CWF material spiked with known quantities of iodine before and after processing. Quantities of iodine measured in each process step have been compared to the predicted, physics-code estimate for radiogenically produced iodine and agree within the measurement and code uncertainties.Future planned measurements for iodine will also be discussed in this presentation. Future work will include direct measurement of iodine in the fuel itself prior to electrochemical treatment, and analysis of the fuel-rod gas plenum. The measurement of iodine in the fuel will allow direct comparison of radiogenic iodine concentrations to the predicted physic-code values; and, while radiogenic iodine is not expected in the gas plenum region of the spent fuel rod, this hypothesis will be tested.
10:45 AM - AA9.5
Coated Magnetic Nanoparticles for Acidic Nuclear Waste Separation.
Maninder Kaur 1 , Jozef Kaczor 2 , Hongmei Han 1 , Andrew Johnson 2 , Andrzej Paszczynski 2 , You Qiang 1
1 Department of Physics and Environmental Science Program, University of Idaho, Moscow, Idaho, United States, 2 Environmental Biotechnology Institute, University of Idaho, Mscow, Idaho, United States
Show AbstractMagnetic Nanoparticles (MNPs) with large specific surface area are promising for separating nuclear High Level Waste (HLW) [1]. Conjugated with actinide specific chelators, the MNP-chelator complex can selectively adsorb the radioactive elements which are then collected magnetically [2]. Due to the highly acidic conditions found in most nuclear waste tanks, the MNPs are soluble in acidic solution. This would cause the MNPs to lose their magnetic property and unable to be used for magnetic nuclear waste separation. Therefore, protective coating of MNPs is necessary to prevent the metal leaching under acidic conditions. To this aim, silica coatings with versatile surface modification and good dispersibility are of particular interest, which need to be optimized for both the excellent acidic resistance and good magnetic susceptibility for the separation. The Fe-Fe3O4 core-shell MNPs used in this study have a mean particle size smaller than 30 nm. The sol-gel coating procedure [3] was employed to coat the MNPs with silica layer. Then, the silica-coated MNPs were incubated in 1M HCl and their acid resistance was evaluated by following the release of soluble iron over time using a colorimetric iron assay. The magnetic hysterisis loops of the silica coated MNPs were measured at room temperature by vibrating sample magnetometer. Their morphology was characterized by transmission electron microscope.The leaching experiment indicates that no detectable Fe3+ ions leach for 25 days when the concentration of the tetraethoxysilane (TEOS) precursor solution used in the reaction reaches to 10%. This suggests that the silica-coating produced by using 10% of TEOS is enough for the required acidic protection. The saturation magnetization (Ms) of the core-shell MNPs drops by increasing TEOS concentration. At 10% of TEOS, the Ms still remains as large as 49 emu/g, which is high enough for magnetic separation. Therefore, a combination of excellent acidic resistance and good magnetic properties is achieved for the silica-coated MNPs at 10% of TEOS. Furthermore, the enhancement of 237Np(V) sorption and extraction is also observed after conjugating with actinide chelators, which indicates that the silica coating on MNPs not only prevents the acid invasion of the MNP, but also is beneficial to nuclear waste separation.[1] L. Nuñez, & M. D. Kaminski. J. of Magnetism and Magnetic Materials 194, 102 (1999).[2] R. D. Ambashta, P. K. Wattal, S. Singh, and D. Bahadur. Separation Science andTechnology 41, 925 (2006).[3] Q. X. Liu, Z. H. Xu, J. A. Finch, and R. Egerton. Chemistry of Materials 10, 3936 (1998).
11:30 AM - AA9.6
On Long Term Spent Fuel Dissolution, Impacts of Surface Precipitation.
Daqing Cui 1
1 Spent Fuel Chemistry, Studsvik AB, Nykoping Sweden
Show AbstractLong term spent fuel dissolution at different repository conditions was systematically investigated. Under air saturated synthetic groundwater solution (Allard water with 2mM NaHCO3), the inventory fractions of radionuclides leached from 2 cm long spent fuel segments (with cladding)with burn-up in the range of 27-75.4 MWD/kg (PWR) were found to be: group (1), low volatile temperature fission products (FP) (I-127/129 > Cs-137 > Rb-85/87) ≈ group (2), redox sensitive FP (Tc-99 > Mo100) ≈ group (3), divalent FP (Ba-138 > Sr-90) > group (4), UO2 matrix, U-238 > group (5), minor actinides Np-237 > Pu-239 > group (6), La group elements Eu > Nd-144 > Ce-140> La-139 ≈ Nd-144 ≈ Pr-141. The radionuclides leaching rates (inventory fractions per day) was found to be dramatically decreased with leaching time, even the synthetic groundwater is under saturated the solubilities of U(VI) phases. The phenomenon is interpreted by the surface precipitation on the spent fuel surface. An experiment was designed and conducted to verify the hypothesized the influences of divalent cations (Ca and Mg) precipitate on UO2.00 (as spent fuel simulator) oxidative dissolution. It was observed that in an air saturated synthetic groundwater solution with 2mM NaHCO3, 0.45mM Ca2+, 0.18mM Mg2+ and 10 mM ion strength, UO2.00 displays four times slower oxidative dissolution rate than that in the solution containing only 2mM NaHCO3, 10 mM Na+ or K+. Microscopic analysis (SEM-EDS and TEM-diffraction) revealed that there was a Ca-Mg rich carbonate precipitate on the surface of UO2.00. It demonstrates that the formation of Ca-Mg carbonate precipitate on UO2.00 surfaces does not necessarily follow the role of solubility products that defined for the bulk solution, and can significantly minimize the oxidative dissolution of UO2.00. The potential influence of Fe(III) precipitate on spent fuel dissolution under deoxygenated condition was also investigated. In an argon purged solution containing 2mM NaHCO3 and 10mM NaCl, the dissolution rate of a 2cm long spent fuel segment with cladding was found to be slower than that at corresponding air saturated conditions, but still at rather significant level due to radiolysis. After being interacted in a FeCO3 saturated solution containing about 0.03 mM Fe2+ for one year, Fe(III) precipitates was observed on the surface and in the cracks of spent fuel pellet. Then, the spent fuel segment was leached again at the same deoxygenated conditions as the formal leaching before interacting in the Fe(II) solution. The leaching rates of all radionuclides from the SNF were found to be decreased dramatically. Fe(III) precipitates on the surface of SNF or in the cracks were explained to play an important role as chemical sorbent, transport barrier, and/or physical shield to minimize the water radiolysis.
11:45 AM - AA9.7
Novel Dual-stimuli Responsive Sequestering Agents in Form of Hydro-gel Beads for Radiological Decontamination of Aqueous Solutions.
Dario Deli 1 , David Crouch 1 , Kathleen Doig 2 , Stephen Yeates 1
1 Organic Materials Innovation Centre, Manchester University, Manchester United Kingdom, 2 Centre for Radiochemistry Research, Manchester University, Manchester United Kingdom
Show AbstractDue to increasing environmental and human health concerns, new materials are required to remedy radioactive contamination and selectively remove heavy metals from contaminated waters. N-aza-crown ethers have proved to selectively complex a wide range of metal ions on the basis of ring sizes (1); their recovery after complexation, however, is often difficult due to their good solubility in different solvents and the stability of the complex is sometimes low (2). Incorporation of these compounds into well-designed polymer supports opens the way to novel sequestering agents for radiological decontamination.We report the synthesis and properties of novel pH and temperature stimuli responsive hydro-gel beads incorporating N-aza crown ethers. Two different materials were prepared employing suspension polymerization technique; poly-(N-isopropylacrylamide-co-acrylic acid) hydro-gel and novel poly-[2-(2-Methoxyethoxy) ethyl methacrylate-co-oligo-(ethylene glycol) methyl ether methacrylate Mn=475], post-functionalised with different N-aza crown ethers in order to fine-tune metal binding selectivity. These gel beads show rapid response to dual environmental stimuli and their size can be finely adjusted by environmental changes of pH and temperature. Swollen states mean high adsorption of water and high surface area with metal ions whereas in the collapsed state the material will release water and the ions not selectively retained by the polymer matrix.Preliminary studies for Ce(I), Co(II), Sr(II) and Ni(II) show that these materials can strongly bind heavy metal ions and therefore they are promising candidates for use as smart scavenging agents for radioactive contamination.1 Arnorld, K. A.;, J. C.; Li, C.; Mallen, J.V.; Nakano, A.; Schall, O. F.; Trafton, J. E.; Tsesarskaja, M.; White, B. D.; Gokel, G. W. Journal of Physical Chemistry A 2006, 110, (50), 13568-13577.2 Tunca, U.; Yagci, Y. Progress in Polymer Science 1994, 19(2), 233-86.
12:00 PM - AA9.8
Performance Indicators Quantifying the Contribution of Safety Functions to the Confinement of Radionuclides in a Geological Repository System.
Eef Weetjens 1 , Jan Marivoet 1 , Suresh Seetharam 1
1 Radiological Impact and Performance Assessments, SCK-CEN, Mol Belgium
Show AbstractThe main indicators for the evaluation of the safety of a geological repository are the dose or risk that are estimated for a representative set of evolution scenarios. In recent years complementary sets of safety and performance indicators have been developed. Whereas safety indicators aim at giving an indication on the level of safety provided by the repository system, performance indicators aim at illustrating how the repository system works. Hitherto, most sets of performance indicators in safety cases of geological repositories are related to the multi-barrier concept; they quantify the contribution of the main engineered and natural barriers to the confinement of the radionuclides within the geological repository system. However, around 1995 the application of the defence-in-depth principle has led to the introduction of safety functions in safety cases of geological disposal. Safety functions are used in various sub domains of the safety case, such as development of repository concepts, explaining the functioning of the repository system to different audiences, deriving representative sets of evolution scenarios, and structuring the safety case. Because of the paramount role played by the multi-safety-functions concept in recently published safety cases, we derived performance indicators that quantify the contribution of the safety functions which actively contribute to the confinement of radionuclides within the repository system. The considered safety functions are containment, limitation of release and retardation. The proposed performance indicators are based on time-integrated fluxes of radionuclides released from the main compartments of the repository system. The product of the three indicators for the safety functions gives the confinement factor of the repository system, i.e. the fraction of the inventory present in the disposed waste that is released from the repository system during the considered time span. The proposed indicators can be applied to individual radionuclides as well as to a weighted sum over all radionuclides. Radiotoxicity, for which the dose factors for ingestion are used as weighting factors, is an example of a weighted sum that can be used to give an overall picture of the functioning of the repository system. The proposed indicators are calculated for a geological repository excavated in a clay formation. The obtained results clearly illustrate the functioning of the repository system by quantifying which safety functions contribute to the confinement of the radionuclides. For solubility limited radionuclides the safety function limitation of release often gives a significant contribution to the confinement, whereas for radionuclides that are strongly sorbed on clay minerals the safety function retardation gives the highest contribution. For the actinide decay series a combination of the safety functions limitation of release and retardation results in an excellent confinement within the repository system.
12:15 PM - AA9.9
Deep Life and Gases in the Outokumpu Deep Borehole: Base Line Information for Nuclear Waste Disposal in Crystalline Rock.
Lasse Ahonen 1 , Ilmo Kukkonen 1 , Anu Kapanen 2 , Mari Nyyssoenen 2 , Merja Itavaara 2
1 , Geological Survey of Finland, Espoo Finland, 2 , VTT Technical Research Centre of Finland, Espoo Finland
Show AbstractSeveral studies have confirmed the existence of deep subterranean biosphere extending far deeper than the planned nuclear waste repositories. The Outokumpu deep borehole (2500 m) in eastern Finland provides an excellent platform for studying the hydrogeochemistry and biogeochemistry of the deep crystalline bedrock. We report the first results obtained using the Outokumpu borehole in studying deep biosphere, dissolved gases and related biogeochemical processes in saline waters of fractured rock. Borehole water was sampled using a tube sampler, with which vertical variations in water and gas compositions, as well as in microbiology, can be detected. Surficial microbiological contamination in sampling was minimized by using aseptic procedures. In addition to tube sampling of water in the borehole, long-term pumping of fracture water from a packed-off borehole section (960-972 m) was carried out. Geochemical characteristics of the fracture water (EC, Eh, pH, dissolved oxygen) were continuously monitored during a four weeks pumping period. Samples were taken for microbiological and molecular biological studies through an aseptic sampling line under anaerobic conditions. Fracture zone at the depth of 965 m discharges saline (Ca-Na-Cl-water, TDS 12 about g/L) methane-rich water to the borehole. Stable isotope data (O-18 and H-2) indicate that the discharging fracture water differs completely from meteoric waters. The saline fluid contains about 0.9 L of dissolved gases in one liter of water. Main components of the gas phase are methane (70-80 vol-%) and nitrogen (20-30 vol-%), while helium, ethane, propane and argon as well as occasionally hydrogen are minor constituents. Dissolved sulfate and sulfide in the deep saline water are below the detection limit (10 mg/L and 0.5 mg/L for sulfate and sulfide, respectively). However, analyzed total dissolved sulfur (50-100 mg/L of S) indicates the presence of some unspecified sulfur species. Moreover, fracture water redox conditions are reducing (sulfidic/methanic), and oxygen concentration is below 10-6 M.Total number of microbes decreased downwards in the borehole from about 10^5 cells/mL hole to about 10^3 – 10^4 cells/mL. Several microbial communities, characterized by considerable diversity were detected by16S rRNA PCR amplification and DGGE analysis. Both bacteria and Archaea were identified by sequencing and databank alignment. Functional gene analysis by dsrB DGGE and cloning of mcrA indicated that sulfate reducers and methanogens are present, but with different populations at different depths. Clustering of microbial communities correlated with changes in water chemistry as a function of depth.
12:30 PM - AA9.10
Determination of Crevice Corrosion Susceptibility of Alloy 22 Using Different Electrochemical Techniques.
Mauricio Rincon Ortiz 1 2 , Martin Rodriguez 1 2 , Ricardo Carranza 1 2 , Raul Rebak 3
1 Materiales, Comisión Nacional de Energía Atómica, San Martín, Buenos Aires, Argentina, 2 Instituto Sabato, UNSAM - CNEA, San Martín, Buenos Aires, Argentina, 3 , GE Global Research, Schenectady, New York, United States
Show AbstractAlloy 22 (UNS N06022) belongs to the Ni-Cr-Mo family and it is highly resistant to general and localized corrosion, but it may suffer crevice corrosion in aggressive environmental conditions, such as high chloride concentration, high applied potential and high temperature. Alloy 22 is one of the candidates to be considered for the outer corrosion-resistant shell of high-level nuclear waste containers. It is assumed that localized corrosion will only occur when the corrosion potential (ECORR) is equal or higher than the crevice corrosion repassivation potential (ER,CREV). This parameter is obtained by different electrochemical techniques using artificially creviced specimens. These techniques include cyclic potentiodynamic polarization (CPP) curves, Tsujikawa-Hisamatsu electrochemical (THE) method or other non-standardized methods. Recently, as a variation of THE method, the PD GS PD technique was introduced.The aim of the present work was to determine reliable critical potentials for crevice corrosion of Alloy 22 in pure chloride solutions at 90C. Conservative methodologies (which include extended potentiostatic steps) were applied for determining protection potentials below which crevice corrosion cannot initiate and propagate. Results from PD GS PD technique were compared with those from these methodologies in order to assess their reliability. Results from the CPP and the THE methods were also considered for comparison. The repassivation potentials from the PD GS PD technique were conservative and reproducible, and they did not depend on the amount of previous crevice corrosion propagation in the studied conditions.