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
Tuesday PM, April 06, 2010
Room 3010 (Moscone West)
9:30 AM - **AA1.1
Applying Insights from Repository Safety Assessments.
Peter Swift 1 Show Abstract
1 Organization 06780, Sandia National Laboratories, Albuquerque, New Mexico, United States
Despite 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 Show Abstract
1 , Argonne National Laboratory, Argonne, Illinois, United States
The 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 Show Abstract
1 , Sandia National Laboratories, Carlsbad, New Mexico, United States
The 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 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.  Giambalvo E., Brush L.H., Xiong Y.-L, (2002), Eos Trans. AGU, 83 (47), Fall Meet. Suppl., Abstract U11B-02, 2002. Novak C.F., (1996), J. Contaminant Hydrology, 21, 297-310. Wang, Y.-F., (1998), “WIPP PA Validation Document for FMT (Version 2.4), Document Version 2.4.” Carlsbad, NM: Sandia National Laboratories.  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 Show Abstract
1 EES-12, Los Alamos National Laboratory, Carlsbad, New Mexico, United States
When 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 . 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) . The WIPP PA also considers that uranium will speciate as U(VI) with a probability of 0.5 in the PA vectors . 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. : 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 Show Abstract
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
We 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 Show Abstract
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
On 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 Show Abstract
1 Department of Physics and Materials Science, Uppsala University, Uppsala Sweden
After 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
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 Show Abstract
1 Department of Chemistry, Northwestern University, Evanston, Illinois, United States, 2 Materials Science Division , Argonne National Laboratory, Argonne, Illinois, United States
New 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 Show Abstract
1 Pyroprocessing Technology, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Idaho State University, Idaho Falls, Idaho, United States
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 Show Abstract
1 , SIA Radon, Moscow Russian Federation, 2 , Savannah River National Laboratory, Aiken, South Carolina, United States
The 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 Show Abstract
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
Considering 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 . 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 . 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 , 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 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. 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. 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 Show Abstract
1 Geological Sciences, University of Manitoba, Winnipeg, Manitoba, Canada
The 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.
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
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 Show Abstract
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
Assessing 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 Show Abstract
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
ZrO2 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 Show Abstract
1 Chemistry, University of Manitoba, Winnipeg, Manitoba, Canada, 2 DEN/DTCD/SECM/LDMC, CEA Valrhô Marcoule, Bagnols/Céze France
Crystalline 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 Show Abstract
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
The 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  and (RuO2, RhO2) oxide phases with acicular or polyhedral shapes  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. 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 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 Show Abstract
1 High Level Waste Plants, Sellafield Ltd, Seascale, Cumbria, United Kingdom, 2 , National Nuclear Laboratory, Seascale, Cumbria, United Kingdom
The 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
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 Show Abstract
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
Reprocessing 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 Show Abstract
1 Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
A 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 Show Abstract
1 Civil Engineering, University of Texas at El Paso, El Paso, Texas, United States
Engineered 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 Show Abstract
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
Oxalic 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
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 Show Abstract
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
Nuclear 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 Show Abstract
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
Molecular 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  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. K. M. Beardmore, N. Gronbech-Jensen, Physical Review E, v.57, pp.7278-7287, 1998 B. Jeon and N. Gronbech-Jensen, Computer Physics Communications, v.180, pp.231-237, 2009 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 Show Abstract
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
The 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 Show Abstract
1 Institute of Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe Germany
Assessment 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  and solid-solution formation of various actinides . 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  and aluminosilicates.  H. Geckeis, Th. Rabung, J. Cont. Hydrol., 102, 2008, 187-195 M. Schmidt et al., Dalton Trans., 2009, 6645 – 6650 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 Show Abstract
1 Institute for Transuranium Elements, European Commission - Joint Research Centre, Eggenstein-Leopoldshafen Germany
In 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 Show Abstract
1 Department of Civil Engineering and Geological Sciences, University of Notre Dame, Notre Dame, Indiana, United States
This 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 Show Abstract
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
There 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
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 Show Abstract
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
Information 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 Show Abstract
1 Department of Chemistry, Lab. of Radiochemistry, University of Helsinki, Helsinki Finland, 2 , Posiva Oy, Eurajoki Finland
In 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 Show Abstract
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
From 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 Show Abstract
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
Vitrification 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 Show Abstract
1 Materials, Imperial College London, London United Kingdom
Mixed 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 Show Abstract
1 , SIA Radon, Moscow Russian Federation, 2 , IGEM RAS, Moscow Russian Federation, 3 , Savannah River National Laboratory, Aiken, South Carolina, United States
In 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 Show Abstract
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
This 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 Show Abstract
1 , Universitat Politecnica de Catalunya, Barcelona Spain, 2 , CTM Centre Tecnologic, Manresa Spain
The 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 Show Abstract
1 , JAEA, Horonobe, Hokkaido Japan, 2 , Suncoh Consultants Co., Ltd., Tokyo Japan
The 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 Show Abstract
1 Horonobe Underground Research Center, Japan Atomic Energy Agency, Horonobe-Cho, Hokkaido, Japan, 2 , Golder Associates, Redmond, Washington, United States
A 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 Show Abstract
1 Chemical Engineering, UPC, Barcelona Spain, 2 Environemental Technology Area, CTM, Manresa Spain, 3 , AMPHOS 21, Barcelona Spain
Cements 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 Show Abstract
1 Engineering Materials, University of Sheffield, Sheffield United Kingdom
Ionsiv 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 Show Abstract
1 Applied Quantum Physics and Nuclear Engineering, Kyushu university, Fukuoka Japan
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 Show Abstract
1 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency, Muramatsu Tokai, Ibaraki, Japan
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 Show Abstract
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
In 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  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.  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. 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).