bluuebunny/crossref_metadata_2025_split huggingface dataset:
Modelling Transport of 14 C-Labelled Natural Organic Matter (NOM) in Boom Clay
Authors: Ionescu Alice , Maes Norbert , Mallants Dirk 
Published: April 2011

AbstractIn Belgium, the Boom Clay formation is considered to be the reference formation for HLW disposal R&D. Assessments to date have shown that the host clay layer is a very efficient barrier for the containment of the disposed radionuclides. Due to absence of significant water movement), diffusion - the dominant transport mechanism, combined with generally high retardation of radionuclides, leads to extremely slow radionuclide migration. However, trivalent lanthanides and actinides form easily complexes with the fulvic and humic acids which occur in Boom Clay and in its interstitial water. Colloidal transport may possibly result in enhanced radionuclide mobility, therefore the mechanisms of colloidal transport must be better understood. Numerical modeling of colloidal facilitated radionuclide transport is regarded an important means for evaluating its importance for long-term safety.The paper presents results from modeling experimental data obtained in the framework of the EC TRANCOM-II project, and addresses the migration behavior of relevant radionuclides in a reducing clay environment, with special emphasis on the role of the Natural Organic Matter (NOM) [1]. Percolation type experiments, using stable 14C-labelled NOM, have been interpreted by means of the numerical code HYDRUS-1D [2]. Tracer solution collected at regular intervals was used for inverse modeling with the HYDRUS-1D numerical code to identify the most likely migration processes and the associated parameters. Typical colloid transport submodels tested included kinetically controlled attachment/detachment and kinetically controlled straining and liberation.
Comparative Analysis of Results From Deterministic Calculations of the Release of Radioactivity From Geological Disposal of Spent Fuel in Crystalline Rocks
Authors: Prvakova Slavka , Amosov Pavel V. , Nilsson Karl-Fredrik 
Published: April 2011

AbstractThis paper presents results from deterministic calculations of radionuclide migration in a deep geological environment. The concept assumes single canister with spent nuclear fuel situated in bentonite (near-field) and surrounded by crystalline host rock (far-field). The results are presented in the form of release rates from the near-field and far-field, which contains also a part of advective release from the system of two single fractures. Additionally, radiological risk is evaluated in the form of dose rates for a water drinking scenario for an exposed population group. The analyses have been done by using the methodologies and computer codes used at the Institute for Energy EC-JRC in the Netherlands and the Mining Institute KSC RAS in Russia.
Evaluating Chemical Toxicity of Surface Disposal of LILW-SL in Belgium
Authors: Mallants D. , Wang L. , Weetjens E. , Cool W. 
Published: April 2011

AbstractONDRAF/NIRAS is developing and evaluating a surface disposal concept for low and intermediate level short-lived radioactive waste (LILW-SL) at Dessel, Belgium. In support of ONDRAF/NIRAS's assignment, SCK•CEN carried out long-term performance assessment calculations for the inorganic non-radioactive components that are present in LILW-SL. This paper summarizes the results obtained from calculations that were done for a heavily engineered surface disposal facility at the nuclear zone of Mol/Dessel. The calculations address the migration of chemotoxic elements from the disposed waste to groundwater.Screening calculations were performed first to decide which non-radioactive components could potentially increase concentrations in groundwater to levels above the groundwater standards. On the basis of very conservative calculations, only 6 out of 41 chemical elements could not be classified as having a negligible impact on man and environment. For each of these six elements (B, Be, Cd, Pb, Sb, and Zn), the source term was characterized in terms of its chemical form (i.e., metal, oxide, or salt), and a macroscopic transport model built that would capture the small-scale dissolution processes relevant to element release from a cementitious waste container. Furthermore, reliable transport parameters in support of the convection-dispersion-retardation (CDR) transport calculations were determined. This included derivation of (1) solubility for a cementitious near field environment based on thermodynamic equilibrium calculations with The Geochemist's Workbench, and (2) distribution coefficients based on a compilation of literature values. Scoping calculations illustrated the effects of transport parameter uncertainty on the rates at which inorganic components in LILW-SL leach to groundwater.
Characterisation of Partial Melting and Solidification of Granite E93/7 by the Acoustic Emission Technique
Authors: Spasova Lyubka M. , Gibb Fergus G.F. , Ojovan Michael I. 
Published: April 2011

AbstractThe acoustic emission (AE) technique was used to detect and characterise the processes associated with generation of stress waves during melting and solidification of granite E93/7 at a pressure of 0.15 GPa. The AE signals recorded as a result of partial melting of the granite at a temperature of 780°C and subsequent solidification during cooling were distinguished from the equipment noise and their parameters used to characterise the AE sources associated with the phase transformations during melting and solidification of the granite. The mechanisms generating AE during granite melting were differentiated by AE signals with their highest peaks in the frequency spectrum at 170 and 268 kHz. The transformation of the liquid into glass during solidification of the partially melted granite generated AE waves in an essentially broad range of frequencies between 100 and 300 kHz. This preliminary work demonstrates the potential of the AE technique for use in applications related to deep borehole disposal of radioactive wastes.
SIMS Study of C, O and N Impurity Contamination for Multi-Crystalline Si Solar Cells
Authors: Wang Larry , Hockett R. S. 
Published: April 2011

AbstractThis paper is a case study of using SIMS to quantitatively measure C, O and N impurity contamination at two sequential commercial process steps: (1) Si feedstock: 7N (modified Siemens) and 5N feedstock (UMG-Si); and (2) multi-crystalline (mc-Si) solar wafers: cut and etched, from directional solidification bricks grown from 7N and 7N/5N (80:20) feedstock. The conclusion of this study is twofold: (a) the primary opportunity to reduce C, O and N contamination in mc-Si solar cells is at the directional solidification process, and (b) the costly specification of highly pure Si feedstock is unnecessary from a C, O and N perspective if a directional solidification process is used.
Effects of Deposition Parameters on the Structure and Photovoltaic Performance of Si Thin Films by Metal Induced Growth
Authors: Mersich Peter T. , Verma Shubhranshu , Anderson Wayne A. , Giese Rossman F. 
Published: April 2011

AbstractA metal-induced growth (MIG) process was employed to deposit thin films of microcrystalline silicon (μc-Si) for solar cell applications. Due to different grain orientations of the crystals, the absorption coefficient of μc-Si is about 10 times higher than the absorption coefficient of single crystalline Si. The properties of the Si film were investigated resulting from variations in several parameters. A range of Ni and Co thicknesses were examined from 7.5 nm to 60 nm including combinations of the two, while the dc sputtering power was stepped up from 150 W to 225 W. The structure of the resulting film was studied using scanning electron microscopy (SEM), energy dispersive x-ray spectroscopy (EDS) and x-ray diffraction (XRD). SEM of the film revealed that 5 hr of Si deposition at 150 W yields a film thickness of 6.5 μm and a maximum grain size of about 0.6 μm. EDS data showed that at the middle of the Si film the atomic percentage of the Si was 99.17%. XRD data showed that the dominant crystal orientation is {220}. To characterize the photovoltaic properties of the μc-Si, Schottky photodiodes were fabricated. Ni alone as the seed layer resulted in ohmic behavior. With Co only, MIG formed a rectifying contact with open-circuit voltage (V∝). The combination of Co layered over Ni formed better thin films and gave a Voc of 0.24 V and short-circuit current density (Jsc) of 5.0 mA/cm2 since the Co prevents Ni contamination of the top of the grown Si layer.
CdTe Films on Mo/Glass Substrates: Preparation and Properties
Authors: Valdna Vello , Grossberg Maarja , Hiie Jaan , Kallavus Urve , Mikli Valdek , Traksmaa Rainer , Viljus Mart 
Published: April 2011

AbstractShort-bandgap group II-VI compound cadmium telluride is widely used for the infrared optics, radiation detectors, and solar cells where p-type CdTe is needed. p-type conductivity of CdTe is mainly caused by the chlorine-based A-centers, and in part, by the less stable copperoxygen complexes. As a rule, CdTe films are recrystallized by the help of a cadmium chloride flux that saturates CdTe with chlorine. In chlorine-saturated CdTe A-centers are converted to isoelectronic complexes that cause resistivity increasement of CdTe up to 9 orders of magnitude. Excess copper and oxygen or group I elements as sodium also deteriorate the p-type conductivity of CdTe like excess chlorine. p-type conductivity of CdTe can be restored e.g. by the vacuum annealing which removes excess chlorine from the film. Unfortunately, treatment that betters ptype conductivity of the CdTe film degrades the junction of the superstrate configuration cells. In this work we investigate possibilities to prepare p-type CdTe films on the molybdenum coated glass substrates. Samples were prepared by the vacuum evaporation and dynamic recrystallization of 6N purity CdTe on the top of Mo-coated glass substrates. Then samples were recrystallized with cadmium chloride flux under tellurium vapour pressure. Results of the test studies on the structure and electronic parameters of samples are presented and discussed.
The Development of Sprayed Backfill Technology
Authors: Jaroslav Pacovsky , Jiří Svoboda , Jiří Št'ástka 
Published: April 2011

AbstractThe Centre of Experimental Geotechnics, CTU in Prague has recently launched a sprayed backfill technology development programme. Such technology will, most probably, be used during the future construction of deep underground radioactive waste repositories.The safe disposal of radioactive waste in deep underground geological repositories assumes the construction of a multi-barrier system the aim of which will be to eliminate the potential penetration of hazardous radionuclides into the biosphere over a time span of several hundred thousand years.With regard to the location in the repository in which the containers will be placed, bentonite will seal the space between the container and the rock mass in the form of a highly compacted bentonite block, while access galleries to the nests will be backfilled with a mixture of natural bentonite and rock. This mixture will be compacted in layers using compacting machinery thus ensuring maximum backfill density during compaction. Following compaction, however, spaces will remain in the gallery vault to which such compacting machines will be unable to gain access. Sprayed clay (sprayed backfill) technology, however, will allow the filling of such voids by injecting backfill so that the material compaction level obtained using this method is at least the same as the backfill compaction level resulting from that using standard compacting machinery (vibration roller, vibration plate compactor etc.).Sprayed backfill technology is based on shotcrete technology which is a method commonly applied today in the construction of underground structures. Sprayed backfill technology development may involve the use of concrete spraying machines; however the parameter requirements for the spraying of bentonite backfill differ in principal from those which apply to shotcrete.The article will provide information on the development of sprayed backfill technology, the machinery employed and the sprayed material as well as research results obtained to date.
Studies on radionuclide transport behaviour–The next generation of in-situ experiments at the Grimsel Test Site
Authors: Martin Andrew James , Blechschmidt Ingo 
Published: April 2011

AbstractTwo recent ongoing major projects at the Grimsel Test Site (GTS) (www.grimsel.com) that were initiated to simulate the long-term behaviour of radionuclides in the repository near-field and the surrounding host rock are presented: the Colloid Formation and Migration (CFM) project, which focuses on colloid generation and migration from a bentonite source doped with radionuclides and the Long-Term Diffusion (LTD) project, which aims at in-situ verification and understanding of the processes that control the long-term diffusion of repository-relevant radionuclides. So far, the CFM project has principally involved: development and implementation of a state-of-the-art sealing concept to control hydraulic gradients in a shear zone to imitate repository-relevant conditions; extensive laboratory studies to examine bentonite erosion and colloid formation in a shear zone; and, development of models to estimate colloid formation and migration. The next stage will be to assess the behavior of bentonite colloids generated from a radionuclide spiked bentonite source-term emplaced into the controlled flow field of the shear zone. This will be coupled with further extensive laboratory studies in order to refine and evaluate the colloid models currently used in performance assessments. The LTD project consists of: a monopole diffusion experiment where weakly sorbing and non-sorbing radionuclides (3H, 22Na, 131I, 134Cs) have been circulating and diffusing into undisturbed rock matrix since June 2007; experiments to characterise pore space geometry, including determination of in-situ porosity with 14C doped MMA resin for comparison with laboratory derived data; a study of natural tracers to elucidate evidence of long-term diffusion processes; and, an investigation of the in-situ matrix diffusion paths in core material from earlier GTS experiments. Future experiments will focus on diffusion processes starting from a water-conducting feature under realistic boundary conditions.
Measurement of HLW glass dissolution/alteration kinetics by using micro-reactor flow-through test method
Authors: Inagaki Y. , Mitsui S. , Makigaki H. , Idemitsu K. , Arima T. , Banba T. , Noshit K. 
Published: April 2011

AbstractA new type of flow-through test method using micro-reactor was developed and applied to measurement of the dissolution/alteration kinetics for a Japanese type of simulated HLW glass, P0798. In this test method, a face of coupon shaped glass specimen (30mm × 10mm × 4mm size) is in contact with a micro-channel (20mm length, 2mm width, 0.16mm depth) constructed on a PTFE (Teflon®) plate, and a solution is injected into the inlet of micro-channel at a constant rate. The injected solution, which flows through the micro-channel reacting with the glass to the outlet, is retrieved at certain intervals to be analyzed for determination of the glass dissolution/alteration rate. After the test, the glass specimen removed from the micro-reactor is subjected to surface analyses. This test method has major features as follows, 1) any controlled solution condition can be provided over the test duration, 2) a relatively high S/V ratio can be provided by use of micro-reactor in spite of using coupon shaped glass specimen, which results in precise and consistent analyses of both the solution and the reacted glass surface, 3) the test apparatus is simple with compact size and easy operation, which allows a flexible setup of test conditions. By use of this test method the dissolution/alteration rate for P0798 glass was measured as a function of pH, temperature, and time, and the results indicated that this test method is applicable and suitable for evaluation of the dissolution/alteration kinetics.
Use of Archaeological Glass to Predict the Long-Term Behavior of HLW
Authors: Gin S. , Verney-Carron A. , Libourel G. 
Published: April 2011

AbstractThe predictability of models describing long-term nuclear glass behavior in a geological repository can be tested by means of natural or archaeological analogs. This study covers fractured archaeological glass blocks from a shipwreck discovered near the Mediterranean island of Embiez (France). The blocks were examined mainly because of their morphological analogy with nuclear glasses. Fractured after production (as in the case of nuclear glass), these blocks had been leached for 1800 years in seawater. The laboratory investigation led to the development and subsequent validation on archaeological objects of a geochemical model capable of accurately simulating the coupling between chemistry and transport to account for the alteration state of the cracks according to their geometric characteristics. Laboratory experiments allowed us to determine the kinetic and thermodynamic parameters for modeling glass alteration. The model was then tested against short-term experiments before simulating the crack alteration over 1800 years. We show that cracks in the outer regions of the block are the most severely altered because of rapid solution renewal, whereas internal cracks are very slightly altered because of a rate-limiting effect of water transport due to the formation of secondary phases. This study also establishes a direct link between data obtained at lab scale and the long-term evolution of a complex system in a natural environment, indicating that the key phenomena have been identified experimentally. The analogous behavior of archaeological and nuclear glass during leaching experiments and the similarities in their crack networks allow us to consider applying the model to nuclear glasses under geological repository conditions. This study clearly shows that the internal crack network does not play a major role in the overall long-term alteration of archeological glass blocks. The issue of the transposition studies will be to determine whether this conclusion can be generalized to nuclear glasses.
Composition Changes and Future Challenges for the Sellafield Waste Vitrification Plant
Authors: Riley A. , Walker S. , Gribble Nick R. 
Published: April 2011

AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors produced during reprocessing of spent nuclear fuel by bonding the fission products as metal oxides into a borosilicate glass matrix. This provides a stable and durable waste form suitable for safe long term storage and ultimate disposal.WVP was commissioned with feed from reprocessing of Magnox reactor fuel. This material is relatively low in fission product content per tonne of fuel, but contains significant Al and Mg from fuel cladding. WVP also routinely treats a blended feed made from a mixture of Magnox and Oxide reprocessing products. The Oxide fuel from Light Water Reactor (LWR) and Advanced Gas Cooled (AGR) power stations is of higher burnup and contains more fission products per tonne of fuel, also Gd and other process additives. Blending allows 25% incorporation of waste oxides by weight in glass to be achieved routinely.Recent programmes of development work in WVP have been aimed at increasing incorporation rates for these feeds, to reduce the number of waste containers produced for disposal. Work has also focussed on increasing the throughput of WVP, to more rapidly treat current stocks of liquid reprocessing waste, both by increasing the feed rate and by improving the lifetime of key components to improve plant availability.Future challenges for WVP include flowsheet changes to treat historic stocks of reprocessing wastes containing high U, Fe and Cr. Washout of solids from the base of waste storage tanks in preparation for decommissioning is also likely to give high Mo feeds. Development of flowsheet and glass formulation to accept these changes in feed composition will be a key objective of future work.
The Impact of Increased Waste Loading on Vitrified HLW Quality and Durability
Authors: Gribble Nick R , Short Rick , Turner Edward , Riley Andrew D 
Published: April 2011

AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquid waste (HAL) arising from the reprocessing of spent nuclear fuel in the UK. In order to optimise WVP operations a full scale working replica of a WVP processing line, the Vitrification Test Rig (VTR), was constructed to processes non-active HAL simulants. Recently the VTR has been used to determine an operational envelope for the vitrification of HAL from Magnox reprocessing at a waste oxide incorporation rate in glass of up to 35wt% (compared to a “standard” incorporation rate of 25wt%). This paper discusses the differences in operating conditions necessary to achieve acceptable waste throughput at the increased incorporation rate. The chemical durability of the resulting vitrified product is also discussed, along with the formation of secondary phases, and a comparison is drawn between 35wt% incorporation glasses and products made at the standard 25wt% incorporation.
Measurement of initial dissolution rate of P0798 simulated HLW glass by using micro-reactor flow-through test method
Authors: Makigaki H. , Inagaki Y. , Idemitsu K. , Arima T. , Mitsui S. , Banba T. , Noshita K. 
Published: April 2011

AbstractWe applied a new type of flow-through test method using micro-reactor consisting of a simple test apparatus with compact size to measurement of the dissolution rate of a Japanese type of simulated waste glass (P0798 glass). In this test method, a solution flows through a micro-channel (20 mm length, 2 mm width, 0.16 mm depth) in contact with a face of coupon shaped glass specimen, and the output solution is retrieved at certain intervals to be analyzed for determination of the glass dissolution rate. By using this test method the initial dissolution rate of glass matrix or forward dissolution rate was measured as a function of pH (3 to 11) and temperature (25°C to 90°C). The present test results indicated that the initial dissolution rate has ‘V-shaped’ pH dependence, and the effect of pH on the dissolution rate decreases with increase in temperature similar to the results measured by using the Single-pass flow-through (SPFT) method. The present test results also indicated that the dissolution of B is controlled by diffusion process and that of Si is controlled by surface reaction process.
Microwave Vitrification of Model Heavy Metals Carriers From Wastewaters Treatment
Authors: Kovacova Milota , Lovas Michal , Jakabsky Stefan , Romero Maximina , Rincon Jesus Ma 
Published: April 2011

AbstractThe metallurgical industry of ferrous and non-ferrous metals produce huge amount of wastes. In Slovakia, the factory for nickel production was closed in 1993, but around 5.5 kt of wastes remain in a dump. This waste was used as a model sorbent of heavy metals (Cu, Cd, Co) from wastewater treatments. The TCLP (Toxicity Characteristic Leaching Procedure) test of precipitated heavy metals on waste materials has confirmed the necessity of waste stabilization. The microwave vitrification was applied because of a high content of iron in waste. After vitrification, the chemical durability and microhardness by indentation have been tested in the glassy and glassceramics obtained.