bluuebunny/crossref_metadata_2025_split huggingface dataset:
Effect of Stainless Steel Can/Glass-ceramic Interaction Layer on Aqueous Durability
Authors: McGlinn Peter , Zhang Yingjie , Li Huijun , Payne Timothy E. 
Published: October 2011

AbstractCalcined high-level radioactive waste (HLW) stored at the Idaho National Laboratory (INL) will eventually be immobilised in a suitable wasteform before disposal. A tailored glass-ceramic wasteform, produced by hot isostatic pressing in stainless steel (SS) cans, has been developed at ANSTO as a cost-saving alternative to glass which would improve waste loading and density, and reduce waste volume. We have studied the SS/wasteform interactions under HIPing conditions to understand whether such interactions would have any detrimental effect on long-term wasteform stability. This has been demonstrated by carrying out aqueous durability tests, under near-neutral and alkaline conditions, on the wasteform at the interaction layer, and on the wasteform distal to this reaction edge. Reaction during HIPing resulted in Cr diffusion from the can wall into the wasteform, however without any detectable detrimental impact on the HIP can or the aqueous durability of the wasteform.
Helium Migration Mechanisms in Polycrystalline Uranium Dioxide
Authors: Martin Guillaume , Desgardin Pierre , Garcia Philippe , Sauvage Thierry , Carlot Gaëlle , Barthe Marie-France 
Published: October 2011

AbstractThis study aims at identifying the release mechanisms of helium in uranium dioxide. Two sets of polycrystalline UO2 sintered samples presenting different microstructures were implanted with 3He ions at concentrations in the region of 0.1 at.%. Changes in helium concentrations were monitored using two Nuclear Reaction Analysis (NRA) techniques based on the 3He(d,α)1H reaction. 3He release is measured in-situ during sample annealing at temperatures ranging between 700°C and 1000°C. Accurate helium depth profiles are generated after each annealing stage. Results that provide data for further understanding helium release mechanisms are discussed. It is found that helium diffusion appears to be enhanced above 900°C in the vicinity of grain boundaries possibly as a result of the presence of defects.
Chlorine Diffusion in Uranium Dioxide : Thermal Effects versus Radiation Enhanced Effects
Authors: Pipon Yves , Toulhoat Nelly , Moncoffre Nathalie , Bererd Nicolas , Jaffrezic Henri , Barthe Marie France , Desgardin Pierre , Raimbault Louis , Scheidegger Andre M. , Carlot Gaelle 
Published: October 2011

AbstractChlorine is present as an impurity in the UO2 nuclear fuel. 35Cl is activated into 36Cl by thermal neutron capture. In case of interim storage or deep geological disposal of the spent fuel, this isotope is known to be able to contribute significantly to the instant release fraction because of its mobile behavior and its long half life (around 300000 years). It is therefore important to understand its migration behavior within the fuel rod. During reactor operation, chlorine diffusion can be due to thermally activated processes or can be favoured by irradiation defects induced by fission fragments or alpha decay. In order to decouple both phenomena, we performed two distinct experiments to study the effects of thermal annealing on the behaviour of chlorine on one hand and the effects of the irradiation with fission products on the other hand. During in reactor processes, part of the 36Cl may be displaced from its original position, due to recoil or to collisions with fission products. In order to study the behavior of the displaced chlorine, 37Cl has been implanted into sintered depleted UO2 pellets (mean grain size around 18 μm). The spatial distribution of the implanted and pristine chlorine has been analyzed by SIMS before and after treatment. Thermal annealing of 37Cl implanted UO2 pellets (implantation fluence of 1013 ions.cm−2) show that it is mobile from temperatures as low as 1273 K (Ea=4.3 eV). The irradiation with fission products (Iodine, E=63.5 MeV) performed at 300 and 510 K, shows that the diffusion of chlorine is enhanced and that a thermally activated contribution is preserved (Ea=0.1 eV). The diffusion coefficients measured at 1473 K and under fission product irradiation at 510 K are similar (D = 3.10-14 cm2.s−1). Considering in first approximation that the diffusion length L can be expressed as a function of the diffusion coefficient D and time t by : L=(Dt)1/2, the diffusion distance after 3 years is L=17 μm. It results that there is a great probability for the chlorine contained in the UO2 grains to have reached the grain boundaries after 3 years, in the core of the fuel rod as well as at its periphery. Moreover, diffusion and concentration of chlorine at grain boundaries has been evidenced using SIMS mapping. Our results indicate therefore, that, during reactor operation and after, the majority of 36Cl is likely to have moved to grain boundaries, rim and gap. This fraction might then significantly contribute to the rapid or instant release of chlorine. This could have important consequences for safety assessment.
Computational Investigation of the Formation of Hyperstoichiometric Uranium Dioxide (UO2+x)
Authors: Skomurski Frances , Becker Udo , Ewing Rodney 
Published: October 2011

AbstractThe stability of spent nuclear fuel as a waste form largely depends upon the oxidation behavior of uranium dioxide (UO2) under repository conditions. Because U6+ phases are more soluble than U4+ phases, and changes in oxidation state can lead to volume changes which increase the susceptibility of spent fuel to corrosion, understanding the mechanisms behind the formation of more highly oxidized UO2 phases is important. While a variety of diffraction and spectroscopic studies have been used to investigate the oxidation behavior of UO2 to U4O9, questions still remain as to the effect of interstitial oxygen on the oxidation state of uranium in hyperstoichiometric UO2. In this study, computational techniques were used to investigate questions such as: Is the formation of U5+ or U6+ more energetically favorable upon the oxidation of UO2A density functional theory (DFT) approach was used to investigate energetically favorable locations of interstitial oxygen atoms and to determine their effect on the charge distribution of uranium in U4O9, specifically. A single unit cell of U4O9 is used as a starting model, and the optimized geometry and charge distribution is tested by using starting models with different U charge assignments. Results from our calculations suggest that the formation of one U5+ per addition of interstitial oxygen at a perpendicular bisector site is favorable. Deflection of two lattice oxygen atoms along the body diagonal of the cubic sites is also observed upon the addition of one interstitial oxygen atom. Structural and bond length data are compared with experimental data whenever possible.
A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles
Authors: Peters M. T. , Ewing R. C. 
Published: October 2011

ABSTRACTThere are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U6+-secondary phases; c) waste form–waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 105 years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms “tailored” to specific geologic settings.
Comparative Study on the Corrosion Resistance of Fe-Based Amorphous Metal, Borated Stainless Steel and Ni-Cr-Mo-Gd Alloy
Authors: Lian Tiangan , Day Daniel , Hailey Phillip , Choi Jor-Shan , Farmer Joseph 
Published: October 2011

ABSTRACTIron-based amorphous alloy Fe49.7Cr17.7Mn1.9Mo7.4W1.6B15.2C3.8Si2.4 was compared to borated stainless steel and Ni-Cr-Mo-Gd alloy on their corrosion resistance in various high-[Cl-] solutions. The melt-spun ribbon of this iron-based amorphous alloy have demonstrated a better corrosion resistance than the bulk borated stainless steel and the bulk Ni-Cr-Mo-Gd alloy, in high-[Cl-] brines at temperatures 90ºC or higher.
Oxide Film Aging on Alloy 22 in Halide Containing Solutions
Authors: Rodríguez Martín A. , Rebak Raúl B. , Carranza Ricardo M. 
Published: October 2011

ABSTRACTPassive and corrosion behaviors of Alloy 22 in chloride and fluoride containing solutions, changing the heat treatment of the alloy, the halide concentration and the pH of the solutions at 90°C, was investigated. The study was implemented using electrochemical techniques, which included open circuit potential monitoring over time, potentiodynamic polarization and electro-chemical impedance spectroscopy (EIS) measurements carried out at open circuit and at passiv-ity potentials. Corrosion rates obtained by EIS measurements after 24 h immersion in naturally aerated solutions were below 0.5 &#61549;m/year. The corrosion rates were practically independent of solution pH, alloy heat treatment and halide ion nature and concentration. EIS low frequency resistance values increased with applied potential in the passive domain and with polarization time in pH 6 - 1M NaCl at 90ºC. This effect was attributed to an increase in the oxide film thick-ness and oxide film aging. High frequency capacitance measurements indicated that passive ox-ide on Alloy 22 presented a double n-type/p-type semiconductor behavior in the passive potential range.
Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials
Authors: Dunn Darrell , Pan Yi-Ming , He Xihua , Yang Lietai , Pabalan Roberto 
Published: October 2011

ABSTRACTThe evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines.
Radiation Tolerance of A2Ti2O7 Materials - A Question of Bonding?
Authors: Whittle Karl , Lumpkin Gregory R , Smith Katherine L , Blackford Mark G , Harvey Elizabeth J , Zaluzec Nestor J 
Published: October 2011

AbstractThe resistance of Ln2Ti2O7 (Ln=lanthanide) compounds to radiation damage is an important topic in the understanding and development of new materials by which radioactive nuclear waste can safely be immobilised. A model has been developed, from previously published density functional theory and molecular orbital theory simulations of the band structure for Ln2Ti2O7 materials. This model provides a chemical interpretation of radiation stability.
Kinetic and Thermodynamic Study of the Chemistry of Neoformed Phases during the Dissolution of Phosphate Based Ceramics
Authors: du Fou de Kerdaniel Erwan , Clavier Nicolas , Dacheux Nicolas , Podor Renaud 
Published: October 2011

AbstractPhosphate matrices are considered as potential candidates for the specific immobilization of tri- and tetravalent actinides. One of the main properties of interest of these matrices concerns their chemical durability. It was studied through their behavior regarding to the retention of actinides and developped using under- and over-saturation experiments. In this field, neoformed phases were precipitated. Lanthanides were used as surrogates for trivalent actinides while uranium and thorium were used for tetravalent actinides. These phases were extensively characterized through several analytical and spectroscopic techniques (SEM, EPMA, XRD, µ - Raman, TRLFS…).New phosphate form such as Nd1-2xCaxThx-yUy(PO4,F). ½ H2O rhabdophane were identified during the dissolution of britholites. The precipitation of Nd1-2xCaxThxPO4. ½ H2O rhabdophane was obtained and it appeared that there was a segregation between neodymium, calcium and thorium after a few days to several months depending on the thorium weight loadings and the precipitation time, thorium precipitating as TPHPH and neodymium as NdPO4 &#61598; ½ H2O. The determination of the solubility constants showed very low values which means that these phases exhibit some interesting properties for the retention of radionuclides.
Pretreatment of Tc-Containing Waste and Its Effect on Tc-99 Leaching from Grouts
Authors: Aloy Albert , Kovarskaya Elena N , Harbour John R. , Langton Christine A. , Holtzscheiter E William 
Published: October 2011

AbstractA salt solution (doped with Tc-99), that simulates the salt waste stream to be processed at the Saltstone Production Facility, was immobilized in grout waste forms with and without (1) ground granulated blast furnace slag and (2) pretreatment with iron salts. The degree of immobilization of Tc-99 was measured through monolithic and crushed grout leaching tests. Although Fe (+2) was shown to be effective in reducing Tc-99 to the +4 state, the strong reducing nature of the blast furnace slag present in the grout formulation dominated the reduction of Tc-99 in the cured grouts. An effective diffusion coefficient of 4.75 x 10-12 (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol. The leaching results show that, even in the presence of a concentrated salt solution, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. The measured diffusivity was introduced into a flow and transport model (PORFLOW) to calculate the release of Tc-99 from a Saltstone Vault as a function of hydraulic conductivity of the matrix.
Fluidized Bed Steam Reformed (FBSR) Mineral Waste Forms: Characterization and Durability Testing
Authors: Jantzen Carol , Lorier Troy H. , Pareizs John M. , Marra James C. 
Published: October 2011

AbstractFluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750°C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.
Rare Earth Elements: A Tool for Understanding the Behaviour of Trivalent Actinides in the Geosphere
Authors: Buil Belen , Gómez Paloma , Garralón Antonio , Turrero M. Jesús 
Published: October 2011

AbstractREE concentrations have been determined in groundwaters, granite and fracture fillings in a restored uranium mine.The granitoids normalized REE patterns of groundwaters show HREE-enrichment and positive Eu anomalies. This suggests that the REE are fractionated during leaching from the source rocks by groundwaters. Preferential leaching of HREE would be consistent with the greater stability of their aqueous complexes compared to those of the LREE, together with the dissolution of certain fracture filling minerals, dissolution/alteration of phyllosilicates and colloidal transport.
Determination of Matrix Diffusion Properties of Granite
Authors: Holtta Pirkko , Siitari-Kauppi Marja , Huittinen Nina , Poteri Antti 
Published: October 2011

AbstractRock–core column experiments were introduced to estimate the diffusion and sorption properties of Kuru Grey granite used in block–scale experiments. The objective was to examine the processes causing retention in solute transport through rock fractures, especially matrix diffusion used to estimate the importance of retention processes during transport in different scales and flow conditions. Rock–core columns were constructed from cores drilled into the fracture and were placed inside tubes to form flow channels in the 0.5 mm gap between the cores and the tube walls. Tracer experiments were performed using uranine, HTO, 36Cl, 131I, 22Na and 85Sr at flow rates of 1–50 µlmin-1. Rock matrix was characterized using 14C–PMMA method, scanning electron microscopy (SEM), energy dispersive X–ray micro analysis (EDX) and the B.E.T. method.Solute mass flux through a column was modelled by applying the assumption of a linear velocity profile and a molecular diffusion. Coupling of the advection and diffusion processes was based on the model of generalised Taylor dispersion in the linear velocity profile. Experiments could be modelled applying a consistent parameterization and transport processes. The results provide evidence that it is possible to investigate matrix diffusion at the laboratory scale. The effects of matrix diffusion were demonstrated on the slightly–sorbing tracer breakthrough curves. Based on scoping calculations matrix diffusion begins to be clearly observable for non–sorbing tracer when the flow rate is 0.1 &#956;l×min-1. The experimental results presented here cannot be transferred directly to the spatial and temporal scales that prevail in an underground repository. However, the knowledge and understanding of transport and retention processes gained from this study is transferable to different scales from laboratory to in–situ conditions.
Surface Complexation of Neptunium(V) with Goethite
Authors: Jerden James L , Kropf A Jeremy 
Published: October 2011

AbstractBatch adsorption experiments in which neptunium bearing solutions were reacted with goethite (alpha-FeOOH) have been performed to study uptake mechanisms in sodium chloride and calcium-bearing sodium silicate solutions. This paper presents results identifying and quantifying the mechanisms by which neptunium is adsorbed as a function of pH and reaction time (aging). Also presented are results from tests in which neptunium is reacted with goethite in the presence of other cations (uranyl and calcium) that may compete with neptunium for sorption sites. The desorption of neptunium from goethite has been studied by resuspending the neptunium-loaded goethite samples in solutions containing no neptunium. Selected reacted sorbent samples were analyzed by x-ray absorption spectroscopy (XAS) to determine the oxidation state and molecular speciation of the adsorbed neptunium. Results have been used to establish the pH adsorption edge of neptunium on goethite in sodium chloride and calcium-bearing sodium silicate solutions. The results indicate that neptunium uptake on goethite reaches 95% at a pH of approximately 7 and begins to decrease at pH values greater than 8.5. Distribution coefficients for neptunium sorption range from less than 1000 (moles/kg)sorbed / (moles/kg)solution at pH less than 5.0 to greater than 10,000 (moles/kg)sorbed / (moles/kg)solution at pH greater than 7.0. Distribution coefficients as high as 100,000 (moles/kg)sorbed / (moles/kg)solution were recorded for the tests done in calcite equilibrated sodium silicate solutions. XAS results show that neptunium complexes with the goethite surface mainly as Np(V) (although Np(IV) is prevalent in some of the longer-duration sorption tests). The neptunium adsorbed to goethite shows Np-O bond length of approximately 1.8 angstroms which is representative of the Np-O axial bond in the neptunyl(V) complex. This neptunyl(V) ion is coordinated to 5 or 6 equatorial oxygens with Np-O bond lengths of 2.45 angstroms. The absence of a clearly recognizable Np-Fe interaction for the sodium chloride sorption tests suggests that neptunium in these solutions adsorbs as an outer-sphere complex. XAS results from the calcium-bearing sodium silicate sorption tests show evidence for a neptunyl(V) inner-sphere surface complex with a Np-Fe interaction at 3.5 angstroms. Desorption tests indicate that samples in which neptunium is bound as inner-sphere complexes show significant sorption hysteresis relative to samples in which neptunium is bound largely as outer-sphere complexes.