bluuebunny/crossref_metadata_2025 huggingface dataset:
NMR Investigation of Cation Distribution in HLW Wasteform Glass
Authors: Holland Diane , Parkinson Ben G , Islam Moinul M , Duddridge Adam , Roderick Jonathan M , Howes Andy P , Scales Charlie R
Published: April 2011

Magic-angle-spinning NMR has been used to establish the structural roles of various cations added to the borosilicate glass which is used for the vitrification of high-level nuclear waste (HLW). Representative surrogate oxides with nominal valencies of +1, +2 and +3 have been studied which span the range of oxides from modifier to intermediate and conditional glassformer. NMR has been carried out on those nuclei which are accessible and the species observed have been correlated with the physical and chemical behaviour. The controlling factor is the manner in which the alkali cations partition between the various network groups, changing the distribution of silicon Qn species and the boron N4 ratio. Identifiable superstructural units are also present in these glasses. The aqueous corrosion rate increases with Q3 content, as does the weight loss due to evaporation from the melt. The activation energy for DC conduction scales with N4. Values of N4 obtained for these glasses deviate significantly from those predicted by the currently accepted model (Dell and Bray) and are strongly affected by the modifier or intermediate nature of the surrogate oxide and also by its effect on the distribution of nonbridging oxygens between the silicate and borate polyhedra.
Overview of the CEA French Research Program on Nuclear Waste
Authors: Vernaz Etienne Y. , Poinssot Christophe
Published: April 2011

This presentation gives an overview of the French major research program on nuclear waste and presents some perspectives in the context of the 2006 Act, while focusing on the main CEA contributions. Development of new conditioning processes combined with in-depth studies of the waste-packages long term behaviour studies allows enhancing the expected performance of waste-packages. Geological disposal performances have been demonstrated based in particular on the relevant understanding of the radionuclides chemistry and migration in complex environment, including tracing field-experiments in ANDRA underground research lab. French research and plans for waste from future reactors will be discussed. It is still a very prospective area as fuel materials that will be used in these reactors is not decided. The capacity to recycle most of the fuel component will be a major point in the choice of an integrated concept including reactor and fuel cycle.
Cold Crucible Vitrification of Uranium-Bearing High Level Waste Surrogate
Authors: Stefanovsky S.V. , Ptashkin A.G. , Knyazev O.A. , Zen'kovskaya M.A. , Marra J.C.
Published: April 2011

Three tests on vitrification of U-bearing DWPF SB2 and SB4 waste surrogates (40-50 wt% waste loading) in a lab-scale (56 mm inner diameter) copper cold crucible with melt surface area of 24.6 cm2 energized from a 5.28 MHz/10 kW generator were conducted. At the feeding of the air-dry batch (3-5 wt% moisture) glass productivity and specific melt productivity achieved under steady-state conditions were 0.64 kg/h and 256 kg/(m2×h) [6150 kg/(m2×day)], respectively. Similar to previously prepared U-free products, the U-bearing vitrified SB2 (highferrous) waste product was composed of U-enriched glassy matrix and high-ferrous spinel structure phase. As expected, the vitreous phase in the vitrified SB2 waste surrogate was enriched with Na, Al, Si, P, S, Ca, Sr, Pb, U oxides and Cl, whereas spinel preferentially accumulated transition metals (iron group elements: Mn, Fe, Ni) as well as Cu and Zn oxides. Iron oxide content in the spinel phase was ∼67–71 wt %. This result was consistent with XRD data and implies that this spinel was close to magnetite with respect to chemical composition and structure. This was consistent with our previous data on characterization of glass crystalline materials containing both U-bearing and U-free SB2 waste surrogate produced in resistive furnace and cold crucibles. The vitrified SB4 (high-alumina) waste products were predominantly amorphous at waste loading of up to 50 wt %. Traces of quartz occurred in the “skull” zone. In all the products, uranium was found to be uniformly distributed within the vitreous phase. Leach rates of Na, Si, Fe, and U from the vitrified SB2 waste surrogate (50 wt % waste loading) determined by a SPFT procedure were found to be 1.38, 1.42, 0.122, and 0.841 (g×m−2×d−1), respectively.
Immobilization of Radioactive Iodine Using AgI Vitrification Technique for the TRU Wastes Disposal: Evaluation of Leaching and Surface Properties
Authors: Sakuragi Tomofumi , Nishimura Tsutomu , Nasu Yuji , Asano Hidekazu , Hoshino Kuniyoshi , Iino Kenji
Published: April 2011

Iodine-129 collected from a reprocessing plant is regarded as the dominant nuclide in terms of safety for TRU wastes disposal in Japan. AgI vitrification (AgI-Ag2O-P2O5) is a potential iodine immobilization technique, which has the advantages of less iodine volatilization (low-temperature vitrification) and high volume reduction efficiency (approx. 1/25 the original waste volume). The iodine immobilization property can be evaluated by examining the surface condition of the AgI glass immersed in water. In this study, immersion tests have been performed on AgI glass in pure water in a 3% H2-N2 atmosphere at room temperature, and the surface characteristics have been examined. The thin layer (<4.3 μm) that is formed has been found to consist of AgI, which may act as a barrier, preventing leaching of glass components. The concepts behind the iodine release model have been proposed based on diffusion and the solubility of the components at the glass surfaces.
Synthesis and Characterisation of Ln2TiO5 Compounds
Authors: Aughterson Robert D. , Lumpkin Gregory R. , Smith Katherine L. , Thorogood Gordon J. , Whittle Karl R.
Published: April 2011

Bulk samples of six Ln2TiO5 compounds with Ln = La, Pr, Nd, Eu, Gd and Tb were prepared and characterised. Most of the samples have a phase purity of ∼95% (based on BEI and EDS) with the predominant secondary phase primarily being Ln2Ti2O7. Using XRD, TEM selected area diffraction and high resolution imaging techniques, we have confirmed the results of previous studies which showed that at room temperature Pr2TiO5, Nd2TiO5, Eu2TiO5 and Tb2TiO5 have orthorhombic structures with Pnma symmetry. The structure of Tb2TiO5 was further monitored as a function of temperature. The relevance of Ln2TiO5 compounds to advanced nuclear fuel cycles is discussed.
Characterisation of 20 Year Old Pu238 -Doped Synroc C
Authors: Hambley M.J. , Dumbill S. , Maddrell E.R. , Scales C.R.
Published: April 2011

In 1987 samples of Pu238 and Cm244 doped Synroc C were prepared in the UKAEA Laboratories at Harwell. They were studied for five years before being archived. During decommissioning of the Harwell laboratories the samples were transferred to Sellafield and the opportunity was taken to conduct further studies. Given the age of the samples, they offer a unique insight into the long term radiation stability of Synroc. To date, the three Pu238 samples have been examined. The alpha decay dose experienced by the samples is estimated to be 3 × 1019 alphas per gram.

The sample allowed to accumulate alpha decay damage (10587) was slightly heterogeneous, with an apparent grain size in the range 3–15νm. EDX analysis confirmed the phases present to be those expected in Synroc C and that the Pu had partitioned predominantly into the perovskite and zirconolite. Microcracking was observed in the hollandite and rutile but cracks arrested once they reached a zirconolite or perovskite grain.

Sample 10588 was annealed after five years, at temperatures up to 1200°C, this sample was microstructurally similar to 10587 at the 20-50νm scale but differs at small scales. A major difference is the presence of both intergranular and intragranular porosity.

Sample 10589 was annealed at the same time as 10588 but differed significantly probably due to actual fabrication temperatures being higher than those recorded.

Information on these samples will be presented, along with a discussion of the implications for the expected long term stability of titanate ceramic wasteforms.
RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary
Authors: Clarens F. , Serrano-Purroy D. , Martínez-Esparza A. , Wegen D. , Gonzalez-Robles E. , Pablo J. de , Casas I , Giménez J. , Christiansen B. , Glatz J.P.
Published: April 2011

The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.

A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.

The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.
Np-Incorporation Into K-boltwoodite
Authors: Shuller Lindsay C. , Ewing Rodney C. , Becker Udo
Published: April 2011

Np-237 (τ1/2 = 2.1 million years) is a potentially important contributor to the total dose for a geologic repository under oxidizing conditions. Further, the Np5+-complexes are mobile aqueous species. Several processes may limit the transport of Np, as well as other actinides: i) the precipitation of Np-solids, ii) the incorporation of Np into secondary uranium phases, and iii) the sorption and reduction of Np-complexes on Fe-oxide surfaces. This study utilizes quantum-mechanical calculations to determine the most energetically favorable Np5+-incorporation mechanisms into uranyl phases, where Np5+-substitution for U6+ requires a charge-balancing mechanism, such as the addition of H+ into the structure. Experimental results suggest that uranyl structures with charged interlayer cations have a greater affinity for Np5+ than uranyl structures without interlayer cations. Therefore, the uranyl silicate phase boltwoodite (KUO2(SiO3OH)(H2O)1.5) is selected for this computational investigation. The charge-balancing mechanisms considered to occur with substitution include: i) addition of H+, ii) substitution of Ca2+ for K+, and iii) substitution of P5+ for Si4+. While the incorporation energy results (1-3 eV)are higher than energies expected based on current experimental studies, solid-solution calculations are used to estimate the limit of Np incorporation for the P5+ substitution mechanism (10 ppm at ̃100°C). The electronic structure of the boltwoodite structure provides insight into the electron density that may be involved in the incorporation of Np into the structure.
Uranium(VI) Uptake by Synthetic Calcium Silicate Hydrates
Authors: Tits J. , Fujita T. , Tsukamoto M. , Wieland E.
Published: April 2011

The immobilization of U(VI) by C-S-H phases under conditions relevant for the cementitious near field of a repository for radioactive waste has been investigated. C-S-H phases have been synthesized using two different procedures: the “direct reaction” method and the “solution reaction” method.

The stabilities of alkaline solutions of U(VI) (presence of precipitates or colloidal material) were studied prior to sorption and co-precipitation tests in order to determine the experimental U(VI) solubility limits. These U(VI) solubility limits were compared with the U(VI) solubilities obtained from thermodynamic speciation calculations assuming the presence of combinations of different solid U(VI) phases. The solid phase controlling U(VI) solubility in the present experiments was found to be CaUO4(s).

The U(VI) uptake kinetics and sorption isotherms on C-S-H phases with different C:S ratios were determined under various chemical conditions; e.g., sorption and co-precipitation experiments and different pH’s. U(VI) was found to sorb fast and very strongly on C-S-H phases with distribution ratios (Rd values) ranging in value between 103 L kg-1 and 106 L kg-1. Both sorption and co-precipitation experiments resulted in Rd values which were very similar, thus indicating that no additional sorption sites for U(VI) were generated in the co-precipitation process. Furthermore, C-S-H synthesis procedures did not have a significant influence on U(VI) uptake. The U(VI) sorption isotherms were found to be non-linear, and further, increasing Ca concentrations resulted in increasing U(VI) uptake. The latter observation suggests that U(VI) uptake is controlled by a solubility-limiting process, while the former observation further indicates that pure Ca-uranate is not the solubility-limiting phase. It is proposed that a solid solution containing Ca and could control U(VI) uptake by C-S-H phases.
PHREEQC Modelling of Leaching of Major Elements and Heavy Metals From Cementitious Waste Forms
Authors: Martens Evelien , Jacques Diederik , Gerven Tom Van , Wang Lian , Mallants Dirk
Published: April 2011

In this study, Ca, Mg, Al, and Pb concentrations leached from uncarbonated and carbonated ordinary Portland cement – dried waste incinerator bottom ash samples during single extraction tests (EN12457 test) at a pH from 1 to 12, were modelled using the geochemical code PHREEQC. A good agreement was found between modelling results and experiments in terms of leached concentrations for Ca, Mg, and Al by defining a single set of pure mineralogical phases for both the uncarbonated and carbonated (three levels) samples. The model also predicted well the observed decrease in Ca leaching with increasing carbonation. Modelling results further revealed that leaching of Pb is not controlled by dissolution/precipitation of pure Pb containing minerals only (carbonates and (hydr)oxides). The addition of solid solutions (calcite-cerrusite and gibbsite-ferrihydrite-litharge solid solutions) and adsorption reactions on amorphous Fe- and Al-oxides improved the model representation of the experimentally observed amphoteric leaching profile of Pb from the cementitious material.
A New Natural Analogue Study of the Interaction of Low-Alkali Cement Leachates and the Bentonite Buffer of a Radioactive Waste Repository
Authors: Alexander W. Russell , Arcilla Carlo A. , McKinley Ian G. , Kawamura Hideki , Takahashi Yoshiaki , Aoki Kaz , Miyoshi Satoru
Published: April 2011

Bentonite plays a significant barrier role in many radioactive waste repository designs, where it has been chosen due to its favourable properties such as plasticity, swelling capacity, colloid filtration, low hydraulic conductivity and its stability in relevant geological environments. However, bentonite is unstable at high pH meaning that it could lose its favourable properties if interacted with hyperalkaline leachates from concrete construction materials (e.g. tunnel liners, grouts, etc.), seals and plugs and/or cementitious wastes in a repository. This fact has forced several national programmes to assess alternative construction and sealing materials such as low alkali cements. Recently, it has been assumed that the lower pH (typically pH 10-11) leachates of such cements will degrade bentonite to a much lesser degree than ‘standard’ OPC-based cement leachates (generally with an initial pH>13).

To date, few laboratory orin situURL (underground rock laboratory) data are available to support the use of low alkali cements in conjunction with bentonites, partly because of the very slow kinetics involved. Consequently, a new project has focussed on finding an appropriate natural analogue site to provide long-term supporting data which will avoid the kinetic constraints of laboratory and URL experiments. Early results have identified an initial, very promising site at Mangatarem in the Philippines, where a quarry excavating bentonite and zeolites is found in the sedimentary carapace of the Zambales ophiolite. In the immediate vicinity of the quarry, ophiolite-derived hyperalkaline groundwaters are present and further field work (including geophysics surveys and borehole drilling) are now being planned to assess regional bentonite/hyperalkaline groundwater interaction. This paper presents an overview of the current status of the project and assesses the relevance of the study to improving understanding of low-alkali cement leachate/bentonite interaction.
Apparent Inversion of the Effect of Alloyed Molybdenum for Corrosion of Ordinary and Enhanced 316L Stainless Steel in Sulfuric Acid
Authors: Kwong Gloria , Carcea Anatolie , Newman Roger C.
Published: April 2011

An aging assessment of the OPG waste resin storage system predicted the potential for premature failure of the carbon steel resin liners. Consequently, resin liners made of 316L stainless steel with a minimum content of 2.5% molybdenum were selected to replace the carbon steel liners. The 2.5% Mo 316L stainless steel was specified to enhance pitting resistance in the spent resin environment. With the additional Mo, one would expect that a brief electrochemical corrosion test will reveal the superiority of such alloy over conventional 316L steel. This study reports a contrary experience
Observation of Diffusion Behavior of Trace Elements in Hardened Cement Pastes by LA-ICP-MS
Authors: Chida Taiji , Sugiyama Daisuke
Published: April 2011

Diffusion coefficients of iodine (I), cesium (Cs) and strontium (Sr) in solid cement blocks of ordinary portland cement (OPC) and low-heat portland cement containing 30 wt% fly ash (FAC) are measured by in-diffusion experiment. By using “laser ablation inductively coupled plasma mass spectrometry”, the penetration profiles of trace elements by diffusion were obtained quantitatively in cement solids. The apparent diffusion coefficients of the near-surface of the cement samples were estimated to be about 10-13 m2 s-1for I, Cs and Sr. Lower diffusion coefficients were observed for FAC than those obtained for OPC.
Effect of thermal annealing on characteristics of polycrystalline silicon used for solar cells
Authors: Gou Xianfang , Li Xudong , xu Ying , Liang Xinqing , Zhao Yuwen
Published: April 2011

Behavior of oxygen, carbon and minority-carrier lifetimes of multi-crystalline silicon (mc-Si) has been investigated by means of FTIR and QSSPCD after three step annealing. For comparison, the annealing of czochralski (CZ) silicon was also carried out under the same conditions. The results revalted that oxygen and carbon concentration pf mc-Si had a larger decrease than that of CZ -Si, which means the more oxygen precipitates in mc-Si were generated. High density defects of mc-Si such as grain boundaries and dislocations accelerated formation of oxygen precipates and defect complex compounds are generated after three step annealing, which could be suction center of defect, that reduced carrier decentralized recombination centers that resulted in improvement of lifetime of wafer. Tendency of difference of lifetime was correlated with interior structure of crystalline silicon.
Improvement of the interface quality between Zn (O,S,OH)x buffer and Cu(InGa)(SeS)2 surface layers to enhance the fill factor over 0.700
Authors: Kushiva K. , Tanaka Y. , Hakuma H. , Goushi Y. , Kijima S. , Aramoto T. , Fujiwara Y. , Tanaka A. , Chiba Y. , Sugimoto H. , Kawaguchi Y. , Kakegawa K.
Published: April 2011

For realizing the proof of mass production capablity or a move toward a GW/a production, 16%-efficiency project has been started setting the target of each parameter as V∝: 0.685 V/cell, FF: 0.735 and Jsc: 31.8mA/cm2. Up to FY2008, the target of each parameter independently has been achieved expect the efficiency. All of our research by adjusting the two resistance (Rsh and Rs) in the monolithically integrated 30cm×30cm-sized circuits. To improve the FF, double buffer structure with CBD-Zn(O,S,OH)x and MOCVD-ZnO is proposed and the thickness is adjusted by optimizing the Rs and the Rsh. As the result, FF of over 0.7 has been achieved for the first time in our CIS R&D since FY 1993.
Light-Induced Passivation of Si by Iodine Ethanol Solution
Authors: Sopori Bhushan , Rupnowski Przemyslaw , Appel Jesse , Guhabiswas Debraj , Anderson-Jackson LaTecia
Published: April 2011

We report on our observations of light-activated passivation (LIP) of Si surfaces by iodine-ethanol (I-E) solution. Based on our experimental results, the mechanism of passivation appears to be related to dissociation of iodine by the photo-carriers injected from the Si wafer into the I-E solution. The ionized iodine (I−) then participates in the formation of a Si-ethoxylate bond that passivates the Si surface. Experiments with a large number of wafers of different material parameters indicate that under normal laboratory conditions, LIP can be observed only in some samples–samples that have moderate minority-carrier lifetime. We explain this observation and also show that wafer cleaning plays an extremely important role in passivation.
Understanding Radionuclide Migration From the D1225 Shaft, Dounreay, Caithness, UK
Authors: Savage David , Watson Claire , Wilson James , Bond Alex , Jones Warren , Metcalfe Richard , Milodowski Tony , Munro Colin , Penfold James , Watson Sarah
Published: April 2011

A 65 m vertical shaft was sunk at Dounreay in the 1950s to build a tunnel for the offshore discharge of radioactive effluent from the various nuclear facilities then under construction. In 1959, the Shaft was licensed as a disposal facility for radioactive wastes and was routinely used for the disposal of ILW until 1970. Despite the operation of a hydraulic containment scheme, some radioactivity is known to have leaked into the surrounding rocks. Detailed logging, together with mineralogical and radiochemical analysis of drillcore has revealed four distinct bedding-parallel zones of contamination. The data show that Sr-90 dominates the bulk beta/gamma contamination signal, whereas Cs-137 and Pu-248/249 are found only to be weakly mobile, leading to very low activities and distinct clustering around the Shaft. The data also suggest that all uranium seen in the geosphere is natural in origin. At the smaller scale, contamination adjacent to fracture surfaces is present within a zone of enhanced porosity created by the dissolution of carbonate cements from the Caithness flagstones during long-term rockwater interactions. Quantitative modelling of radionuclide migration, using the multiphysics computer code QPAC shows the importance of different sorption mechanisms and different mineralogical substrates in the Caithnesss flagstones in controlling radionuclide migration.
The methodology followed in Belgium to investigate the compatibility with geological disposal of Eurobitum bituminized intermediate-level radioactive waste
Authors: Valcke Elie , Marien An , Geet Maarten Van
Published: April 2011

In Belgium, Eurobitum intermediate-level long-lived bituminized radioactive waste containing large amounts of NaNO3, which is a hygroscopic and soluble salt, is to be disposed of in an underground repository in a geologically stable clay formation. The Boom Clay is studied as a potential host formation because of its favourable properties to limit and delay the migration of the leached radionuclides and other contaminants (heavy metals, NaNO3, organic molecules) to the biosphere. The emplacement of the bituminized waste will induce multiple processes that could have a significant effect on the key properties of the clay. Because several of these processes are interdependent, the study of the compatibility of Eurobitum with geological disposal is complex. To structure the research and to identify possible knowledge gaps, the Belgian Radioactive Waste Management Agency ONDRAF/NIRAS developed a new methodology based on safety functions and safety statements. In this paper, this methodology is briefly explained, with reference to the disposal of Eurobitum. Experimental results obtained at the Belgian Nuclear Research Centre SCK•CEN are presented and discussed in the light of the safety functions and safety statements approach. The importance of the interdependence of the processes is highlighted. Special attention is given to the evolution of the disposal design as a result of the improved understanding of key processes.
Estimation of seismic observation records in the vicinity of large earthquakes and its impact on geological disposal program
Authors: Tsuchi Hiroyuki , Oouchi Taishi , Ichikawa Yoshikazu , Okutsu Kazuo , Sasaki Toru
Published: April 2011

Japan is located at the converging plate boundaries and is one of the most earthquake-prone zones in the world. In order to ensure the safety of a geological repository against earthquakes, a site with a high possibility of direct destruction by active faults should be excluded, and the relationship between the characteristics of seismic ground motion, subsurface structures, geological disposal system and the propagation characteristics of seismic ground motion should be fully investigated. Earthquake-resistant design based on the latest technology is also very important for ensuring the safety of a geological repository.

Following rapid improvement of seismic observation networks after the Hyogoken Nanbu earthquake in 1995, numerous seismic observation records have been obtained in the vicinity of large earthquakes. According to these seismic observations, some phenomena that might affect the safety of a geological disposal system have occurred. Some earthquakes occurred in the areas where active faults had not been identified, while some records showed that seismic motion in the deep underground environment was greater than that at the surface. We have identified the implications from the latest information concerning large earthquakes for the geological disposal program. This study made it clear that detailed investigation incorporating state-of-the-art technologies could reduce the likelihood of missing active faults to an extremely low level and a more practical analysis of seismic ground motion could be achieved by taking the latest information into account.
French Industrial Vitrification Plant: 30 Years Old and Still Innovating
Authors: Chauvin E. , Ladirat C. , Quang R. Do
Published: April 2011

In 2008, AREVA NC Industrial Vitrification of High-Level Liquid Waste blows out its 30th candle, with always two main objectives during all the time: containment of the long lived fission products and reduction of the final volume of waste. During all this time AREVA with the French Atomic Energy Commission (CEA) developed and use in their industrial installations a selection of borosilicate glass that have been demonstrated as the most suitable containment matrix for waste from spent nuclear fuel. Consistent and long-term R&D programs associated to industrial feed back from operation have enabled continuous improvement of the process: throughput and waste loading factor enhancement. The Vitrification Process used and currently implemented in the AREVA facilities will be described.
Iron Aluminium Silico-Phosphate Glasses Including U3O8
Authors: Oliver Carlos J.R. Gonzalez , Russo Diego O. , Lovey Francisco C. , Romero Maximina , Rincon Jesus Ma.
Published: April 2011

The obtention of iron aluminium silicate- phosphate glasses including U3O8 oxide has been carried out and the study of crystalline phases formation. A serie of these original glasses have been investigated by thermal cycles at 780? C for promoting the nucleation and growth of crystalline phases of iron aluminium phosphates including U3O8 in the 5.1- 13.71 wt% range. Thermal expansion and DTA/G analysis have been carried out showing behaviour trend to crystallization of these glasses. The glass with higher uranium oxide content shows liquid phase in phase separation which change of microstructure during HREM observations. It seems produced a pulsed order- disorder processing which takes place under the electron beam. This effect first observed in TEM-HREM on phase separation droplets is discussed.
Quantifying Conservatism of Performance Assessment Calculations by Sorption Model Reduction: Case Study on Near Field Cs Migration in Callovo-Oxfordian Clay
Authors: Weetjens Eef , Martens Evelien , Jacques Diederik
Published: April 2011

Implementation of full thermodynamic models in performance assessment (PA) calculations (large domain and very large timescales) is practically unfeasible due to excessive computational times. The complex competitive sorption processes of radionuclides are often simplified.

In this paper, such simple sorption models (i.e. linear Kd and Langmuir isotherms) were compared with more complex thermodynamic models for a reduced geometry and relatively short timescales (compared to traditional PA calculations).

Within the context of a safety case the value of this study is twofold. Primarily, it provides support information on how to choose adequate parameter values in a consistent simplified analysis for compliance with safety criteria. For the cases studied in this paper it became clear that a linear model is sufficient to represent sorption, but a proper choice of the Kd values is critical. Secondly, the comparison of results from these compliance calculations with more realistic analyses demonstrates quantitatively the safety margin implicitly present in PA calculations through model abstraction.
Mathematical Modelling of Transport Phenomena in Concrete Porous Media
Authors: Plecas Ilija B. , Dimovic Slavko D.
Published: April 2011

Two fundamental concerns must be addressed when attempting to isolate low-level waste in a disposal facility on land. The first concern is isolating the waste from water, or hydrologic isolation. The second is preventing movement of the radionuclides out of the disposal facility, or radionuclide migration. Particularly, we have investigated here the latter modified scenario.To assess the safety for disposal of radioactive waste-concrete composition, the leakage of 137Cs from a waste composite into a surrounding fluid has been studied. Leakage tests were carried out by original method, developed in Vinca Institute [1,2,3,4,5]. Transport phenomena involved in the leaching of a radioactive material from a cement composite matrix are investigated using three methods based on theoretical equations [6,10]. These are: the diffusion equation for a plane source an equation for diffusion coupled to a first-order equation, and an empirical method employing a polynomial equation‥ The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choises for radioactive waste packaging for a future Serbian radioactive waste disposal center.
Characterisation of Granite Fractures From the In-Situ FEBEX Experiment (Grimsel, Switzerland): ossible Effects on Bentonite Colloid and Radionuclide Transport
Authors: Alonso Ursula , Missana Tiziana , Garcia-Gutierrez Miguel , Patelli Alessandro , Albarran Nairoby , Lopez-Torrubia Trinidad , Ceccato Daniele , Rigato Valentino
Published: April 2011

The FEBEX in-situ experiment, installed in 1997 at the Grimsel Test Site (GTS, Switzerland) 400 m depth under the Swiss Alps, simulates a high level radioactive waste repository (HLWR) emplaced in granite. Its initial aim was to study the performance of a bentonite engineered barrier but recently, two new boreholes were drilled in the granite to study the possible bentonite colloid formation and their migration in the granite.

This study presents the characterization performed, at the micrometer scale, of the threemain water conductive fractures that were identified on the granite cores extracted from the newboreholes. These fractures are possible pathways for bentonite colloid transport (or retention),may be source of natural colloids and may condition colloid stability. The nuclear ion beamtechniques µ-Particle X-Ray Emission (µPIXE) and Rutherford Backscattering Spectrometry(RBS) were applied for visualizing and quantifying the elemental composition of the fracturessurface and of the surrounding micro-fractures, as support of the bentonite colloid analyses.
Anionic and Cationic Effects on the Crevice Corrosion Susceptibility of Alloy 22
Authors: Carranza Ricardo M. , Rebak Raul B.
Published: April 2011

Alloy 22 is highly resistant to all forms of corrosion; however, it may suffer crevice corrosion in presence of chloride ions especially at temperatures higher than ambient and at anodic potentials. The susceptibility of Alloy 22 to suffer crevice corrosion is highly dependent on the type of electrolyte solution that is in contact with the alloy including variety of species in solution and their relative concentration. Laboratory research has shown that at a constant chloride concentration, the susceptibility of Alloy 22 to crevice corrosion is not influenced by the nature of the cations present in solution. On the other hand, that nature of the anions is highly influential. Of the anions present in ground water, only chloride is detrimental and the others are inhibitors or innocuous for crevice corrosion susceptibility. That is, the presence of the other anions counter balances the negative effect of chloride. The inhibition effect is explained and the likelihood that Alloy 22 would suffer crevice corrosion in contact with ground water is discussed.
Basic Research Needs on Physical Chemistry of Radionuclides in the Nuclear Fuel Cycle
Authors: Advocat T. , Ferry C. , Goutelard F. , Lamouroux C. , Wagner J.F. , Vautrin-Ul S. , Chaussey A.
Published: April 2011

From uranium-ore treatment to spent fuel recycling, waste treatment and conditioning and to final storage of waste packages, radionuclides are involved in numerous chemical and physical reactions. The understanding of their chemical forms (speciation) and behavior as well (retention, complexation,...), as a function of the environment conditions (T, P, solid/liquid/gas interfaces), are key issues for the development of the industrial nuclear activities. Dedicated analytical tools are needed to determine the radionuclide concentrations and speciation in the liquids, solids and gas, over a wide range of concentrations and matrixes. The obtained experimental data on radionuclide speciation are integrated in dedicated data bases, supporting various models used to simulate the system behavior (i.e. RN migration under geological disposal, RN contamination in the primary fluids of nuclear power plants, RN behavior in the PUREX process, etc.). There are several needs in the following domains of the fuel cycle :

♦ The development of innovative methods to enhance analytical performances of isotopic composition of elements in irradiated fuels or waste streams arising from processed spent fuels. Isobaric interferences may be suppressed by specific ion-molecules reactions in collision/cell coupled with Mass Spectrometer, instead of preliminary chromatographic separations.

♦ The thermo chemistry at high temperature and pressure of the coolant fluids of the nuclear power plants, to model the solid/liquids interactions controlling its contamination by the activated products and hideout processes.

♦ The development of scientific and operational models of the radiolysis of organic molecules and materials, under extreme conditions (γ and α radiolysis), to understand the controlling long-term degradation phenomena ( i.e. H2 degassing in the waste packages).

♦ The fundamental understanding of sorption processes of redox sensitive elements such as U on specific mineral surfaces, in the presence of organic molecules, to develop dedicated tools for radionuclide monitoring and measurement in the environment.
The role of uranium peroxide studtite on the retention of Cs, Sr and Se(VI)
Authors: Gimenez Javier , Sureda Rosa , Pablo Joan de , Casas Ignasi , Martinez-Llado Xavier , Rovira Miquel , Martinez-Esparza Aurora
Published: April 2011

The formation of uranyl secondary solid phases onto the spent nuclear fuel surface might influence the radionuclide concentration in solution via, among others, sorption processes. In this work, the incorporation of some radionuclides onto the uranium peroxide studtite, UO2O2·4H2O, has been tested.

The study was carried out in batch experiments where a known amount of studtite (0.05 g) was put in contact with 20 cm3 of radionuclide solution. Once equilibrium was reached, radionuclide concentrations in solution were determined by ICP-MS. The radionuclide amount attached to the solid was calculated from the mass balance. The S/V values of the experiments were also determined from BET specific solid surface area measurements.

In this work, data on sorption of caesium, strontium, and selenium as a function of pH are presented. The behaviour of caesium and strontium are similar: a relatively high amount of radionuclide is sorbed at neutral to alkaline pH while there is almost no sorption at acidic pH. On the other hand, in the case of selenium, the sorption maximum occurs at acidic pH and there is almost no sorption at alkaline pH. The different behavior of the radionuclides is related to the element speciation in solution and the surface charge of the solid. Strontium and caesium are sorbed at alkaline pH because they are positively charged in solution and the surface of the studtite is negatively charged (>O- groups) while selenium(VI) sorbs at acidic pH because the surface of the studtite is positively charged, and the predominant selenium(VI) species in solution is anionic.

These preliminary data indicate that the sorption capacity of uranyl secondary solid phases such as studtite is an important process to be considered when establishing the migration of different radionuclides released from spent nuclear fuel.
The United States Department of Energy's Environmental Management Program
Authors: Gerdes Kurt D. , Ross Steven L.
Published: April 2011

The year 2009 marks 20 years since the Environmental Management program was first established in the Department of Energy. At that time, nearly 50 years of nuclear activity had left a legacy that included nuclear waste and environmental contamination at more than 100 sites across the United States. The extent of the risk to our citizens and communities was unknown, and certainly many of the processes and technologies to reduce that risk had not yet been invented. Since then, the Department of Energy has closed 86 of 108 sites originally assigned to the program nationwide. The Department of Energy has packaged and safely stored the nation’s entire excess plutonium inventory. The Department has pioneered new technologies that have allowed progress in retrieving millions of liters of tank waste and safely disposing of tens of thousands of cubic meters of transuranic waste. In Fiscal years 2006 and 2007 alone, the Department of Energy demolished approximately 500 buildings (nuclear, radioactive, and industrial) as part of our decontamination and decommissioning projects. Finally, there have been great strides in restoring groundwater contaminated with radionuclides using innovative treatment systems. In August 2005, a rigorous project management system was instituted. This Department of Energy program was built on the principle of prioritizing risk reduction supported by our four guiding tenets of safety, performance, clean-up, and closure. The mission activities at our clean-up sites are targeted at our highest risk activities. In planning its environmental clean-up efforts and developing the budget for those activities, the Department seeks to focus on work that will produce the greatest environmental benefit and the largest amount of risk reduction.
Application of Formal Knowledge Engineering Approaches to Develop a Design Catalogue for a Japanese HLW Repository
Authors: Matsumoto Masaaki , Sato Osamu , Okubo Hiroo , Makino Hitoshi , Ochi Yasuhiro , Hioki Kazumasa , Umeki Hiroyuki , Masuda Sumio , McKinley Ian G.
Published: April 2011

The volunteer siting of disposal facilities for vitrified high-level radioactive waste (HLW) and TRU waste in Japan results in a particular need for flexibility to allow repositories to be tailored to specific sites. Instead of a single reference concept, a “catalogue” of designs for individual repository components is being developed, which can then be combined to form optimum solutions for particular site boundary conditions. As highlighted in a companion paper – Makino et al: Supporting Development of Practical Designs for a Japanese HLW Repository –, which puts the repository design process in the context of the Japanese HLW programme, the complexity of this work justifies use of advanced Knowledge Engineering approaches. In this paper three components are described – development of the knowledge base, assessing tacit system understanding and production of innovative solutions to resolve conflicts between requirements.
Maximum Energy that can be Harvested from a Dielectric Elastomer Generator
Authors: Koh Adrian
Published: April 2011

Mechanical energy can be converted to electrical energy using a dielectric elastomer generator (DEG). The maximum amount of energy that can be harvested from a DEG is constrained by various modes of failure and operational limits. Known limiting mechanisms include electrical breakdown, electromechanical instability, loss of tension and rupture by stretch. These limits define a cycle where maximum energy can be harvested. The cycle was represented on work-conjugate planes, which can be used as a guide for the design of practical cycles. The amount of energy harvested is larger when a DEG is subject to equal-biaxial stretching.
The Effects of Precipitates and Mn-bearing Particles on the High Strain-Rate Compression of High Strength Aluminum
Authors: Elkhodary Khalil Ibrahim , Lee William , Cheeseman Bryan , Sun Lipeng , Brenner Donald W , Zikry Mohammed
Published: April 2011

Transmission electron microscopy (TEM) and specialized dislocation-density based crystal-plasticity formulations and finite-element schemes were used to investigate the effects of nano-sized precipitates and micro-sized Mn-bearing particles on the behavior of Al-Cu-Mg-Ag alloys. By accurately representing crystallography and the morphology of the different precipitates, accurate predictions can be obtained that indicate that the nano-sized Ω and θ’ precipitates promote ductility and toughness by inhibiting shear-strain localization; whereas the micro-sized dispersed particles intensify localization. Collectively, the precipitates and dispersed particles, however, promote the strengthening of the alloy.
Cathodic Protection Effect on the Assessment of SCC Susceptibility of X52 Pipeline Steel
Authors: Contreras A. , Sosa E. , Espinosa-Medina M. A.
Published: April 2011

Assessment of anodic and cathodic potentials on stress corrosion cracking (SCC) of API X52 pipeline steel through slow strain rate tests (SSRT) was studied. The SSRT were carried out in a NS4 solution to simulated dilute ground water that has been found to be associated with SCC of pipelines. SSRT were performed and evaluated in air and in the NS4 solution at room temperature at an extension rate of 1×10-6 in/sec. Tests were performed at controlled electrochemical polarization potentials, both anodic and cathodic (100, 200, 400 mV) versus the open circuit corrosion potential. The results of reduction in area ratio (RAR), time to failure ratio (TFR) and plastic elongation ratio (PER) of the specimens tested in the soil solution indicate that X52 pipeline steel was susceptible to SCC at cathodic potentials. These specimens showed a brittle type of fracture with transgranular appearance. The SCC proceess and mechanism of X52 steel in the NS4 solution is mixed-controlled by both anodic dissolution and the hydrogen involvement. At positive potentials the SCC is based mainly on the anodic dissolution mechanism. When the applied potentials shifted negatively, the SCC on the steel follows mainly hydrogen embrittlement mechanism. This mechanism was confirmed through the internal cracks observed in the specimens.
Brazing Process to Repair Wide Gap Cracks of Inconel 738 Superalloy Components
Authors: Guzmán Isidro , Garza Alejandro , García Felipe , Castillo Jesús
Published: April 2011

Brazing process is a cost effective technique to repair wide gap cracks in turbine components made from difficult to weld nickel base superalloys. In this process boron and silicon are used as melting point depressants, however, form hard and brittle intermetallic compounds with nickel (eutectic phases) which are detrimental to the mechanical properties of brazed joints. In this paper the effect of brazing parameters such as temperature and time on final microstructure of brazed joint of nickel base superalloy Inconel 738 using a commercial filler metal alloy (Ni-11Cr-3.5Si-2.25B-3.5Fe) was investigated. The microstructure of the joint layer was characterized by optical and scanning electron microscopy; chemical composition was carried out by energy dispersive X-ray spectrometry (EDS) microanalysis and microhardness testing. The results showed that the formation of eutectic microconstituents, within the joint regions, was significantly influenced by the brazing parameters and gap size, also that formation of eutectic constituents decreased by allowing a sufficient amount of time for a complete isothermal solidification to take place at the brazing temperature.
Optical Characterization of the Twisted Nematic Liquid Crystal Display
Authors: Rickenstorff Carolina , Ostrovsky A. S.
Published: April 2011

A simple optical characterization technique is developed for to find the amplitude and phase modulation modes for the twisted nematic liquid crystal display (TN-LCD). The developed technique is based on intensity and pixel shift measurements of the interference pattern obtained from the TN-LCD sandwiched between two polarizers in a modified Young's experiment. Particularly, specimen LC2002 Holoeye is characterized and its amplitude and phase modulation curves as a function of image's gray level, bright B and contrast C are presented.
Synthesis of Au/TiO2 Thin Films for Photocatalytic Degradation of Methylene Blue
Authors: Palomar F. , Gómez I. , Zambrano Patricia
Published: April 2011

Au/TiO2 thin films have been prepared from Titanium Tetrabuthoxide-acetylacetone solution and buthanol as solvent by sol-gel and dip-coating technique. Au nanoparticles were prepared from HAuCl4 solution. The deposition of Au nanoparticles was by spray pyrolysis method. The films from the sol gel solution were heat treated at 450°C for 1 hour. The surface structures, morphology, composition and optical properties on the films were investigated by atomic force microscopy (AFM), optical microscopy (OM), X-ray Diffraction (XRD) and UV-Vis spectrometer. It was found the film consisted of anatase phase (TiO2), Au nanoparticles in the range of 50 to 100 nm, and few cracks on the surface remaining attached to the glass substrate. In this study, we measured the photocatalytic degradation of a methylene blue (MB) aqueous solution by a film prepared by the sol-gel and dip-coating methods TiO2 and Au/TiO2films prepared by spray pyrolysis method. We found that Au nanoparticles deposited on the TiO2films improved the photocatalytic activity and minor degradation time that TiO2 films.