bluuebunny/crossref_metadata_2025 huggingface dataset:
Discussion on the Use of Matrix Diffusion Model after a Multidisciplinary Study of a Granitic Boulder Sample
Authors: Marcos N. , Siitari-Kauppi M. , Suksi J. , Rasilainen K. , Finch R. , Hellmuth K. -H.
Published: May 2011

This paper aims to discuss the applicability of the classical matrix diffusion model against the integrated body of new data obtained by different methodologies on several samples of three granite boulders. The matrix diffusion model was tested against observations from the upper (most weathered in contact with air) and lower (fresh in contact with the ground) part of a boulder block. A U(VI) enrichment up to nearly 300 ppm (compared to about 10 ppm background concentration) mostly as uranophane was observed in the zone between the weathered and fresh rock. U-series disequilibrium studies indicated that most of U has been accumulated recently, about 10 000 years ago [1].

High interconnected porosity (total porosity of > 1% and up to about 5.5% in altered minerals) characterizes the weathered zone (upper part), whereas the maximum porosity values in the fresh zone (lower part) of the rock are about 0.4 – 0.6%. Stable isotope studies δ18O and δ2H confirm that the mineralogical changes observed in the weathered upper part are due to old hydrothermal events. That is, the alteration is much older than uranium accumulation. Mössbauer spectroscopy showed that the Fe(III) content of the biotites from the upper to the lower part decreases from 30% in the weathered zone to 17% in the fresh rock, thus indicating possible redox control for the observed U precipitation. Fission track studies showed that secondary U(VI) also occurs within minerals grains (especially plagioclase) in the upper part.

Mathematical simulations indicate that matrix diffusion alone is not enough to reconstruct the past U accumulation. The simulated concentrations derived from U concentration in pore water multiplied by Kd are clearly too small, indicating apparent insufficiency of the Kd approach. However, even with only matrix diffusion, the simulations roughly reconstruct the observation that U levels are clearly higher in the upper part of the boulder than in the lower part.
Brownian Simulation of Matrix Diffusion
Authors: Andersson Sofie , Emrén Allan T.
Published: May 2011

The commonly used approach in dealing with matrix diffusion is to assign an effective diffusion constant for the radionuclide in the rock matrix. The idea behind this approach is that, on a scale much larger than the pore size, the irregularities tend to cancel out. Although it might look plausible at first sight, this approach has been questioned both for theoretical and experimental reasons.

Here, Brownian simulation has been used to investigate the transport of dissolved material in a rock matrix modeled as a system of pores with a wide variability in size and shape. The Boltzmann distribution is used locally, although the system globally is far from equilibrium.

The simulation consists of two main parts. First, the model rock is formed by precipitation of irregular mineral grains from a liquid phase. As the grains grow, they tend to form a mostly solid piece of rock.

In the second part of the simulation, a dissolved species is introduced at one side of the rock and allowed to diffuse through its pore system. It is found that no apparent diffusion constant, D, can explain the properties of the system. Rather, D is found to be a function of both distance and time.
Radionuclide Movement during Unsteady, Unsaturated Soil Water Flow
Authors: Smiles D.E. , McOrist G.D.
Published: May 2011

Absorption of aqueous solution by soil appears to result in piston-like displacement of the soil solution originally present by the invading solution. Diffusive redistribution of non-reactive solutes, relative to the water, occurs about that front when the solutions are of different chemical constitution. Adsorption of solute by the soil solid retards the solute front relative to the piston- front; it also reduces the effective diffusivity of the solute relative to the water. Such behavior tests prediction of radionuclide transport and retardation during non-steady water flow in unsaturated soil based on measured distribution coefficients.

Experiments, where initially uniform soil columns absorbed tritiated water containing 60Co2+ from a source at constant water potential, illustrate these effects. Columns were sampled at significantly different elapsed times and the transient water content, water soluble salts, tritium and cobalt profiles determined. Theory anticipates, and experiment confirms, self- similarity of these profiles when data is graphed in terms of distance divided by the square root of time. Water, and soil solid-based, material coordinates facilitate data analysis; measurement of absorption isotherms is also discussed.
Migration Mechanisms of Americium(III) through Soil Layers
Authors: Tanaka T. , Ogawa H. , Muraoka S.
Published: May 2011

Migration experiments of 241Am(III) have been performed using a column system, to investigate migration behavior of 241Am through a column packed with porous sedimentary materials: a coastal sandy soil and a reddish soil. Sorption mechanisms of 241Am on the sedimentary materials were examined by a chemical extraction method.

In the case of the reddish soil, most of 241Am sorbed on the influent edge of the column. The 241Am sorbed on the reddish soil was mainly controlled by a reversible ion exchange reaction. The migration of 241Am in the reddish soil could be evaluated by using the Kd concept. In the case of the sandy soil, on the other hand, considerable amount of particulate 241Am passed through the column. The 241Am sorbed on the sandy soil was controlled by irreversible reactions. The migration behavior of particulate 241Am in the sandy soil could be expressed by filtration theory.
Matrix Diffusion Measurements – Through Diffusion versus Electrical Conductivity Measurements
Authors: Ohlsson Y. , Neretnieks I.
Published: May 2011

Matrix diffusion laboratory experiments in dense porous rock are generally very time consuming and one is limited to rather short diffusion lengths, as well as to a small amount of samples. The large heterogeneity of rock, on the other hand, demands a large quantity of samples that are large enough to exclude effects from e.g. increases in interconnected porosity compared to that of the pristine rock.

Electrical conductivity measurements are very fast and larger samples can be used than is practical in ordinary diffusion experiments. The effective diffusivity of a non-charged molecule is readily evaluated from the measurements, and influences from surface conductivity on diffusion of cations can be studied.

In this study traditional through diffusion experiments as well as electrical conductivity measurements are carried out on the same rock samples. The formation factor is determined by both methods, and the methods are compared and discussed.

The surface conductivity is studied by exchanging the surface sites with Na+, Sr2+ and Cs+. After leaching out the free pore ions the surface conductivity is measured.

With the electrical conductivity method the formation factor is determined directly, whereas it has to be calculated using the bulk liquid diffusion coefficient in the diffusion experiments. This causes some uncertainties in the comparison between the experiments. In estimating the bulk liquid diffusivity, the value for infinitely diluted solutions and in pure water environment is commonly used. The calculated formation factor may therefore be somewhat underestimated.
Formation Factor Determinations byIn-SituResistivity Logging
Authors: Löfgren M. , Ohlsson Y. , Neretnieks I.
Published: May 2011

Matrix diffusion in bedrock is traditionally studied by laboratory liquid diffusion experiments, which are time consuming and expensive. A new way of studying matrix diffusion is to measure the electrical resistivity of the rock. This could be done either in laboratories orin-situ. A fast method of obtaining a formation factor log, later used in matrix diffusion calculations, for an entire borehole is proposed.

It is a standard procedure in geophysical well logging to measure rock resistivity and there are well-developed tools for this in the oil industry. The SKB, Sweden, uses boreholes with a small diameter (56 mm) and this reduces the options in choosing resistivity tools. Therefore they have so far relied on the Normal log that only gives quantitative measurements in special cases, after corrections are made. Modern tools, such as the slimhole Dual-Laterolog, are accurate with a high vertical resolution.

The pore fluid resistivity is required when obtaining a formation factor log. In previous work the borehole fluid resistivity has been used. A new method measures the resistivity in groundwater from local fractures. These values seem more appropriate to use as newin-situmeasurements show that there may be local groundwater resistivity differences.

A preliminary study shows that the conductivity could be obtained in saline Swedish groundwater by measuring the chloride concentration only.

Anomalies in the formation factor log have to be sorted out by using supporting core logging and non-electricalin-situmethods.
In-Situ Interaction between Cement and Clay: Implications for Geological Disposal
Authors: Sneyers A. , Paul M. , Tyrer M. , Glasser F.P. , Fays J. , Iseghem P. Van
Published: May 2011

The extent and the consequences of interactions between cementitious materials used in radioactive waste management and clay host rock are described. In-situ tests were performed on seven cement formulations representing materials applied in repository construction, for backfilling or for solidification of radioactive waste. Samples were exposed to realistic repository conditions of the Boom Clay Formation in the HADES underground laboratory. Chemical, physical and mineralogical changes across the cement-clay interface were identified by combined observations from Electron Probe Microanalysis, Infrared microscopy and X-Ray powder diffraction. Significant interactions in both the cement and the clay part were found in a zone extending up to several hundreds of microns. The most prominent features are (1) leaching of cement with loss of calcium and/or silicon; (2) development of a calcium-rich zone in Boom Clay close to or at contact; (3) the formation of a contact zone marked by the precipitation of a (hydrated) magnesium aluminate phase; (4) reduction in apparent porosity of initially porous/permeable materials and (5) precipitation of calcite within the cement. This elemental exchange tends to diminish pH and reduce the buffering capacity of the cement. Although hydroxide will diffuse into the clay, the development of an extensive alkaline halo in the surrounding clay is unlikely owing to the buffering capacity of the Boom Clay pore water.
Development and Characterization of Borosilicate Glasses for Immobilization of Plutonium-containing Alkaline Sludges
Authors: Aloy A.S. , Iskhakova O.A. , Koltsova T.I. , Trofimenko A.V. , Jardine L.J.
Published: May 2011

Several glasses were melted using laboratory-scale microwave (MW) equipment to study the vitrification of Pu-containing sludges from the Mining and Chemical Combine in Zheleznogorsk, Russia. This work is part of an effort to identify glass compositions that can be adopted for industrial scale production. The glass must meet durability requirements and have good technological properties for long-term storage and geologic disposal.

Electron Probe Analysis (EMPA), scanning electron microscope (SEM) and x-ray methods were used to characterize glasses containing 50 wt% solid sludge residues with high sodium content.

Leach tests of 3, 7, 14, and 28 day durations were performed using the MCC-1 method at 90°C. The surface of the glass was analyzed before and after leaching. Based on the observed results, there are differences in the behavior of Pu and U for the glasses tested, and glass durability is dependent on homogeneity, the uranium oxidizing state, and the SiO2/Na2O+Li2O ratio.
Protective Effect of the Alteration Gel: A Key Mechanism in the Long-Term Behavior of Nuclear Waste Glass
Authors: Gin S.
Published: May 2011

Experimental alteration results are presented in which a French SON 68 (R7T7-type) nuclear containment glass specimen was first altered under static conditions for 600 days at an S/V ratio of 5000 m−1 before transferring it to pure water. The experiment was designed to assess the diffusion barrier properties of the gel formed during the preliminary alteration phase. Contrary to predictions by kinetic models based only on chemical affinity, the renewed alteration of the specimen in pure water was very limited. Measurements at close intervals showed that the maximum alteration rate under these conditions was 7 ×10−3 g·m−2d−1, or about r0/300, whereas a pristine glass coupon in contact with pure water is altered at r0.

The renewed alteration behavior is attributed to partial (5–10%) dissolution of the existing gel, resulting in a slight degradation of its protective properties. Saturation conditions, at a different level than in the preliminary phase, were reached within a few days. The saturation of the solution with respect to silicon, generally interpreted as a glass/solution (Grambow) or gel/solution (Bourcier) equilibrium, is shown to depend not only on the alteration conditions, but also on the specimen leaching history, and is thus not an inherent glass property.

This experiment confirms the idea that the gel formed under saturation conditions controls the kinetics of SON 68 glass alteration by means of a diffusion barrier effect. It also raises a number of issues concerning the concept of a “residual rate” and the possible relations between the quasi steady-state Si concentrations observed in solution and the protective properties of the gel.
Monte Carlo Modelling of Glass Dissolution: Comparison with Experiments
Authors: Lobanova M. , Maurer L. , Barboux P. , Devreux F. , Minet Y.
Published: May 2011

One presents the results of numerical simulations of glass leaching. The glass is modelled as a random mixture of partly and totally soluble species, which represent silica, and boron or alkali oxides, respectively. It is shown that the dissolution rate and the thickness of the altered surface layer are strongly dependent on the glass composition, whereas the equilibrium solubility is not. The dependence of the layer thickness on the glass surface area to solution volume ratio is also emphasized. The protective role of the surface layer is shown to arise from its restructuring after the extraction of the soluble species. The simulation results are compared to an experimental study performed on series of SiO2-B2O3-Na2O glasses.
Nd- and Ce-doped ceramic-glass composites: chemical durability under aqueous conditions and Surface alteration in a moist clay medium at 90°C
Authors: Mcglinn P.J. , Advocat T. , Loi E. , Leturcq G. , Mestre J.P.
Published: May 2011

Zirconolite-glass and sphene -glass specimens, doped with REE as simulants for trivalent actinides, were leached under two conditions. The first was a solubility test using a powdered sample in deionized water at an SA/V ratio of 200 cm−1, to examine the long-term leaching behavior of the composite materials. The other test was carried out in the presence of moist clay, to assess the degree of surface alteration of the composites in the presence of potential geological repository materials. Both tests were carried out at 90°C.
High Level Liquid Waste Solidification Using a “Cold” Crucible Induction Melter
Authors: Demine Andrei V. , Krylova Nina V. , Polyektov Pavel P. , Shestoperov Igor N. , Smelova Tatyana V. , Gorn Valerii F. , Medvedev Gennadii M.
Published: May 2011

At the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a “cold” crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter.

This paper describes the features of a process and equipment for a two-stage HLLW solidification technology using a “cold” crucible induction melter (CCIM) and identifies the advantages compared to a one-stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporator and cold-crucible induction melter currently operational at the IA.Mayak. facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak – first to produce borosilicate glass waste forms and then mineral-like waste forms.
Development and Characteristics of a New Porous Glass Crystalline Matrix (Gubka) for Stabilizing Radioactive and Hazardous Solutions
Authors: Anshits A. G. , Aloy A. S. , Tretyakov A. A. , Knecht D. A. , Tranter T. J. , Sharonova O. M. , Vereshchagina T. A. , Zykova I. D. , Burdin M. V. , Sapozhnikova N.V. , Strelnikov A. V. , Macheret J.
Published: May 2011

This paper describes the results of a joint research program of Russian institutes at St. Petersburg, Krasnoyarsk and Zheleznogorsk and the Idaho National Engineering and Environmental Laboratory to develop a new material for stabilizing radioactive solutions and other uses, such as high-temperature catalysis. An open-cell glass crystalline porous material, Gubka, or “sponge” in Russian, having an open-cell porosity of up to 90 %, was produced from hollow glass crystalline microspheres (cenospheres) formed in fly ash from coal combustion. The cenospheres were separated into fractions based on grain size, density, magnetic properties, and whether or not they were perforated. Selected fractions were molded and agglomerated by sintering with or without a binder at high temperatures. Depending on the cenosphere fractions selected, sintering conditions and additional treatments, Gubka was formed with an open-cell porosity ranging from 40-90 %. The porous material has a bulk density of 0.3-0.6 g/cm3, and two types of porous openings: 0.1-30 micrometer flow-through pores in the cenosphere walls and 20-100 micrometer interglobular pores between the cenospheres. Examples of Gubka application described in this paper include stabilization of different surrogate radioactive waste solutions containing 0.0001 to 0.7 M nitric/hydrochloric acid and 0 to 1.2 M sodium nitrate. Waste solid loadings of 46-55 wt.% nitrate salts, or 26-37 wt.% oxides after calcination, were achieved in those tests.
Perovskite Ceramics from Mechanically Activated Batches for Immobilization of Rare Earth– Actinide Fraction of HLW
Authors: Chizhevskaya S.V. , Cherniavskaya N.E. , Ochkin A.V. , Chekmarev A.M. , Stefanovsky S.V.
Published: May 2011

The effects of different methods of grinding or mechanical milling of precursor mixtures were determined for the structure and properties of perovskite ceramics with the formulation Ca1-xGdxTi1-xAlxO3 (0 ≤x ≤ 1). The ceramics were prepared by coldpressing finely ground precursor mixtures at 100-300 MPa to form pellets that were then sintered at 1300-1500oC. Ceramic samples prepared from precursor material batches treated in a ball or planetary mill or ground in a mortar contained unreacted oxides and had low mechanical integrity and chemical durability. The ceramics produced from precursor material treated by high performance mechanical milling using a high speed rotating vortex layer of ferromagnetic bodies were composed of a major perovskite-structured phase (90-95%) with a Ca-Ti-Gd-Al-O composition and a minor pyrochlore-structured phase with a composition close to that of gadolinium titanate; these ceramics had the highest bending strength (200-340 MPa) and density (~90% of theoretical) and the lowest leach rates for uranium, plutonium, and americium among all the samples studied.
Leaching Behavior and α-Decay Damage Accumulation of UO2 Containing Short-Lived Actinides
Authors: Rondinella V.V. , Cobos J. , Matzke Hj. , Wiss T. , Carbol P. , Solatie D.
Published: May 2011

UO2 containing short-lived α-emitters, the so-called α-doped UO2, can simulate type (i.e. α- decay) and level of activity of spent fuel at the time when it might become exposed to groundwater in a geologic repository during storage. This allows studying α-radiolysis effects on the dissolution of the fuel matrix. Additionally, UO2 with high concentrations of α-emitters accumulate, during experimentally acceptable short times, the amount of decays, hence of property modifications, corresponding to long storage times for spent fuel. UO2 containing ∼10 and ∼0.1 wt% 238Pu was fabricated and tested. Leaching experiments in deionized water under unaerated conditions, with continuous monitoring of the evolution of the redox potential and pH in the leaching solutions, were performed. The Eh measurements showed a fast increase of the redox potential in the case of the material with the highest α-activity, while the UO2 containing ∼0.1 wt% 238Pu increased its potential more slowly. The redox potential for undoped UO2 decreased steadily during the experiment. As previously observed, higher fractions of U were released in the case of α-doped UO2 compared to undoped UO2. The fractions of U and Pu released during leaching from the α-doped materials were very similar, suggesting that congruent dissolution occurred. After leaching times longer than 10 h, only dissolved species were present in the solutions. Under these experimental conditions, characterized by relatively low values of the ratio sample surface/leachant volume, a dependence of the released amounts on the α-activity of the samples was observed. Periodical measurements of parameters like hardness, showed a rapid buildup of radiation damage in the material with the high α-activity. After more than two years, noticeable changes, namely an increase of the hardness, have begun to be observed also for the material with the low concentration of 238Pu.
Study of BiPbO2NO3 for I-129 Fixation under Reducing Conditions
Authors: Amaya Takayuki , Mukunoki Atsushi , Shibuya Mamoru , Kodama Hiroshi
Published: May 2011

Leaching of the iodide ion from BiPbO2I (BPI), BPI encapsulated in cement (BPIC) and AgI was studied in a low salinity solution and in a high salinity solution under reducing conditions. Although BPI released a limited amount of iodide ions (less than 1%) into the low salinity solution, it released more than 30% of iodide ions into the high salinity solution within 80 days. AgI released more than 30% of iodide ions into both low and high salinity solutions within 80 days. It was proved that BPI is more stable than AgI in the low salinity solution under reducing conditions. BPIC released a limited number of iodide ions (less than 5%) into both low and high salinity solutions. BPIC showed the best leach resistance in the high salinity solution.

BiPbO2NO3 (BPN) was developed to remove iodide ions in a solution and fix them in BPI by the ion exchange reaction. Ion exchange properties under reducing conditions were studied. An anion exchange capacity of 1.0 mEq/g and a distribution coefficient of larger than 0.1 m3/kg were obtained in a solution at a pH of between 9 and 13. The advantages of the process using BPN for removing and immobilizing Iodine-129 were discussed from the standpoint of process simplification.
Corrosion Processes Affecting the Performance of Alloy 22 as a High-Level Radioactive Waste Container Material
Authors: Cragnolino G.A. , Dunn D.S. , Pan Y.-M. , Pensado O.
Published: May 2011

Alloy 22 is the material preferred by the U.S. Department of Energy for the waste package outer container for geological disposal of high-level radioactive waste at the proposed site in Yucca Mountain, Nevada. Alloy 22 is considered to be extremely resistant to various modes of aqueous corrosion over broad ranges of temperature, pH, and concentration of anionic and oxidizing species. Uniform corrosion under passive dissolution conditions, localized corrosion in the form of crevice corrosion, and stress corrosion cracking are discussed on the basis of experimental results obtained with mill annealed, thermally treated, and welded specimens using electrochemical techniques. The approach developed for long-term performance prediction, including the use of empirically derived parameters for assessing localized corrosion and the modeling of the passive dissolution behavior, is described.
Material for the Engineered Barrier System under Development for the LLW Repository in Italy
Authors: Risoluti P. , Coppola Luigi , Collepardi M.
Published: May 2011

In Italy a considerable amount of radioactive wastes has been produced during the about thirty years of nuclear activity. In 1996 began activities for the design of a LLW repository and for a site selection. A near surface LLW Repository based on vault concept is considered, for a national total inventory of around 100,000 cubic meters of conditioned wastes, mostly coming from dismantling of the phased out power stations. A near surface LLW Repository based on vault concept is considered.

The design activity under way includes the Engineered Barrier System (EBS) development and qualification. A cementitious material as backfilling barrier is being considered for waste isolation in a disposal module which would then be emplaced in the disposal unit. The construction material of the module is also under evaluation and qualification.

Possible interaction of the barrier with the chemical environment of the Italian waste inventory has also been identified. This paper describes some of the experimental activity under way for the mix-design identification for the cementitious barrier, along with the results of the first campaign of material characterization.

Different concretes and grouts were investigated with the main aim of assessing the long-term durability of these materials. Preliminary results indicate that crack-free and long-term durable concretes and grouts (300 years) can be manufactured even if exposed to aggressive environments. For the sake of brevity, in this paper the following characteristics only of the concretes will be considered:

- compressive, flexural and tensile strength

- carbonation tests

- chloride diffusion.
GMT - A Large Scale In-Situ Test of the Gas Migration Properties of Engineered Barriers
Authors: Vomvoris S. , Marschal P. , Kickmaier W. , Ando K. , Fukaya M. , Fujiwara A. , Kaku K.
Published: May 2011

The GMT experiment (Figure 1) was initiated in the summer of 1997 under the auspices of RWMC (Radioactive Waste Management Center, Japan). It is part of the experiments conducted within Phase V (1997-2002) of Nagra's Underground Rock Laboratory at Grimsel, in the Swiss Alps [1]. The currents status and results from the site characterization activities and laboratory studies are presented. Future plans are also outlined.
Pore Water Chemistry of the Febex Bentonite
Authors: Fernández A.M. , Cuevas J. , Rivas P.
Published: May 2011

The knowledge of pore water chemistry in the clay barrier is essential for performance assessment purposes in a nuclear waste repository, since the pore water composition controls the processes involved in the release and transport of the radionuclides. The methodology followed to define the representative composition of the FEBEX bentonite pore water is presented in this paper.

A series of bentonite-water interaction tests have been performed with the aim of providing a database on the main chemical parameters of the bentonite. These tests were carried out both with high solid to liquid (s:l) ratios (squeezing tests) and low s:l ratios (aqueous extracts tests). The exchangeable cations have also been analyzed to determine the selectivity coefficient of the exchange reactions. To complete the data set, a physical and mineralogical characterization of the bentonite was made.

The most significant bentonite-water interaction processes controlling the chemistry of the system was identified. The ion concentrations basically depend on the s:l ratio of the system, and the pore water composition is controlled by the dissolution of chlorides, dissolution/precipitation of carbonates and sulphates and the cation exchange reactions in the smectite.

The bentonite/water system was modelled with the PHREEQC2 program to obtain the best possible estimation of the pore water composition for initial conditions of water content (≍14%), after checking the conceptual model with the experimental results. The model predictions fitted satisfactorily with the experimental data at low s:l ratios. At high s:l ratios, the modelled results agree adequately, except for the sulphate content, which could be affected by the effective porosity, anion exclusion or stagnant zones not taken into account in the model. According to the model, the FEBEX bentonite pore water at 14% moisture is a sodium-chloride type, with an ionic strength of 0.25 M and pH of 7.78.
Diffusion Behavior of Ca2+ Ions in Compacted Na-Montmorillonite
Authors: Kozaki Tamotsu , Adachi Yoshifusa , Inada Koichi , Sato Seichi , Ohashi Hiroshi
Published: May 2011

Diffusion behavior of Ca2+ ions in water-saturated, compacted Na-montmorillonite was studied for the safety assessment of geological disposal of high-level radioactive waste. The diffusion coefficients of Ca2+ ions in compacted Na-montmorillonite obtained in this study were from 1.7×10-11 to 6.0×10-12 m2 s-1 with increasing dry density of 1.0 to 1.8 Mg m-3. These values were approximately a quarter those of Na+ ions at each dry density. This suggests that the diffusion of Ca2+ ions could be the rate-determining mechanism for the alteration of Na-montmorillonite into Ca-montmorillonite. The activation energy for diffusion of Ca2+ ions was almost equal to that in free water (17.3 kJ mol-1) at dry densities of 1.0 to 1.6 Mg m-3. However, the activation energy suddenly increased to 25.1 kJmol-1 at a dry density of 1.8 Mg m-3. This increase cannot be explained using a single-diffusion process model, such as the pore-water diffusion model. Diffusion behavior of Ca2+ ions in compacted montmorillonite along with the experimental data on the basal spacing and water content of montmorillonite, and the activation energy for diffusion of Ca2+ and several other ions are discussed.
The Effect of the Excavation Damaged Zone Adjacent to the Walls of Deposition Holes on the Migration of Radionuclides
Authors: Autio J. , Hautojärvi A. , Salo J-P.
Published: May 2011

The excavation damaged rock zone (EDZ) adjacent to the surface of deposition holes in a nuclear waste repository has been considered to be a potential pathway for the flow of water and the migration of radionuclides diffusing out of a waste canister via the bentonite barrier. The properties of the excavation-damaged rock zone adjacent to the surface of experimental deposition holes in the Research Tunnel at Olkiluoto on the southwest coast of Finland have been measured and are used in this study to evaluate the effect of the excavation damaged rock zone on the transport of radionuclides escaping from a waste canister. Since the hydraulic conductivity of compacted bentonite is low, the predominant mode of migration of nuclides through such material will be diffusion. The effect of the excavation-damaged rock zone on the transport of radionuclides was analyzed by comparing two different cases: 1) diffusion through the bentonite and 2) transport through the excavation damaged rock zone by the processes of advection and diffusion. According to this study, hydraulic gradients three orders of magnitude higher than those estimated to currently exist would be required to achieve the transfer of significant quantities of non-sorbing neutral nuclides through the excavation-damaged rock zone. According to this evaluation, the excavation damaged rock zone is unlikely to be a significant migration route in normal cases in the absence of large hydraulic gradients.
Key Aspects of the H12 Safety Case
Authors: Umeki Hiroyuki
Published: May 2011

In Japan, as outlined in the overall high-level radioactive waste (HLW) management program defined by the Japanese Atomic Energy Commission (AEC, 1994), HLW from reprocessing of spent nuclear fuel will be immobilized in a glass matrix and stored for a period of 30 to 50 years to allow cooling. It will then be disposed of in a deep geological formation. Pursuant to the overall HLW management program, an organization with responsibility for implementing HLW disposal will be established around the year 2000. This will be followed by site selection and characterization, demonstration of disposal technology, establishment of the necessary legal infrastructure, relevant licensing applications and repository construction, with the objective of starting repository operation by the 2030s and no later than the mid 2040s.

The HLW disposal program is currently in the research and development (R&D) phase and the Japan Nuclear Cycle Development Institute (JNC) has been assigned as the leading organization responsible for R&D activities. The aim of the R&D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan and to promote understanding of the safety concept not only in the scientific and technical community but also by the general public. One of the features of the R&D program is that its progress is documented at appropriate intervals, with a view to clearly determining the level of achievement of the program and to promote understanding and acceptance of the geological disposal strategy by the general public. As a major milestone, the Power Reactor and Nuclear Fuel Development Corporation (PNC, now JNC) submitted a first progress report, referred to as H3 (PNC, 1992), in September 1992.
Moving HLW-EBS Concepts into the 21st Century
Authors: McKinley I.G. , Kawamura H. , Tsuchi H.
Published: May 2011

Most national high-level waste (HLW) disposal programs actually reflect, or are based on, concepts which were developed during the '70s or early '80s. Although suitable for demonstration of concept feasibility, designs of the engineered barrier system (EBS) do not take into account the tremendous developments in system understanding and materials technology over the last two decades, the practicality (and cost) of their quality assurance and implementation on an industrial scale and the transparency of the demonstration of the safety case. In many ways, due to the increased significance of popular acceptance over the last decade, the last point may be of particular relevance.

This paper reviews the work already carried out on “2nd generation” concepts and extends this to identify the key attributes of an ideal design for the specific case of disposal of vitrified HLW from reprocessing in a “wet” host rock (either crystalline or sedimentary). Based on the concept developed, key R&D requirements are identified.
SR-97: Canister Performance under Normal Disposal Conditions
Authors: Werme Lars
Published: May 2011

The SR-97 disposal canister consists of an inner load bearing cast iron component and an outer corrosion barrier of copper. The mechanical strength of the insert must be such that it withstands, with sufficient safety margins, the combined forces of hydrostatic pressure at the disposal depth (7 MPa) and the bentonite swelling pressure (7 MPa). It must also withstand an increased hydrostatic pressure during a future glaciation. This pressure increase may be as high as 30 MPa. For this load situation, no extra safety margins will be required. Uneven pressure may occur during the water saturation of the buffer and to some extent some uneven pressure may also persist after water saturation. Calculations have been performed for such design basis uneven loads and the present canister design has been shown to withstand them. Under the oxygen-free conditions at the disposal depth, copper will be thermodynamically stable. Only dissolved sulfide will be able to corrode the copper. The supply of sulfide is expected to be low and, consequently, the service life of the canister will be very long. Calculations indicate that no known corrosion processes will lead to canister failure during a shorter time than at least 1 million years.
Approaches to Comparative Risk Assessments of National and International Radioactive Waste Disposal Options
Authors: Heathcote J A , Rintoul A H , Waite P J
Published: May 2011

Pangea Resources International AG, Switzerland, has put forward the concept of an international geological repository in an Australian desert area, for final disposal of certain of the world's high-level radioactive wastes and surplus fissile materials. Already publicly contentious in Australia, the concept raises unusual issues in safety and environmental assessment. It is timely, therefore, to consider why and how an initial case for such a repository might be made, in comparison with national options for high-level waste disposal that it might displace.

There are several stages ahead of development of a formal safety case for an international repository. The first, and almost certainly the most challenging, is winning of public and political acceptance in the host country that such a repository is the ‘best’ solution to a global problem.

We consider the basic need to identify and encompass the concerns of widely disparate stakeholders — industry, governments, national and international regulators, environmental interest groups and the public at large — so that public and political debates can be informed effectively. Many key issues will require comparison both of risks arising from very different operations, and of dissimilar prospective safety performances of complete disposal systems over periods spanning thousands of generations. Nevertheless, we conclude that the validity of such a comparative assessment could be assured by consistent application of a judicious blend of assessment techniques across the alternatives. We also conclude that its usefulness as a vehicle for public discussion would be enhanced by careful attention to public concerns, and by transparently independent review by scientific, technical, sociological and ethical specialists.
Effects of Confining Pressure on Gas and Water Permeabilities of Rocks
Authors: Zhang M. , Takeda M. , Esaki T. , Takahashi M. , Endo H.
Published: May 2011

Knowledge of the permeability of hydraulically-tight rock at great depth is crucially important for the design and/or assessment of facilities associated with underground disposal of radioactive nuclear wastes. This paper presents a recently developed laboratory permeability test system capable of testing low permeability rocks either by using air as a permeant or by the transient-pulse method under high confining pressure conditions that simulate ground pressures at depths. The new system was used to test Shirahama sandstone and Inada granite, which are two types of rock widely available in Japan. To investigate the effects of heterogeneity on rock permeability, specimens cored parallel to and perpendicular to bedding for sandstone, and specimens cored in the direction perpendicular to Rift Plane, Grain Plane and Hardway Plane for granite, were used. The results of this study showed that: 1) gas permeabilities of a dried rock specimen tested by air permeation are almost the same values as water permeabilities of the same saturated rock specimen tested by the transient-pulse method; 2) the intrinsic permeabilities of Shirahama sandstone and Inada granite range from about 8.33E-16 to 7.38E-17 m2 and from 1.86E-17 to 6.94E-20 m2, respectively. They decrease monotonously with increase in effective confining pressure (defined as the difference between the confining and pore pressures), while the rate of decrease diminishes at higher confining pressures. The reduction in permeabilities is due to the closure of microcracks that control fluid flow at low confining pressures; and 3) Inada granite is a heterogeneous and isotropic material. Its hydraulic heterogeneity is more significant in Rift Plane than in Hardway and Grain Planes.
Retrievability; Rationale, Measures, Impacts
Authors: McCombie C. , Zuidema P.
Published: May 2011

The concept of deep geological disposal was developed in order to permanently remove radioactive wastes from the human environment. Retrieval of wastes was reckoned to be a scenario of such low probability that little effort was initially devoted to its study. In recent years, however, there has been an increasingly active debate on whether fully passive (and safe) systems provide a sufficient level of practicability of retrievability. This debate is linked directly to practical, technical matters, such as the design of the facility, the operating procedures and also the institutional programs throughout the lifetime of a repository. But there are also philosophical issues involved in addition to these purely technical issues.

For retrievability, the questions are: How easy does retrieval have to be in the different stages of repository development? What is the rationale for requiring retrievability? What technical measures and methods are feasible? Should specific features facilitating retrievability be introduced into the repository design? How do such measures impact on other aspects of system performance and on other issues (such as safeguards)? This paper addresses the questions raised and presents a subjective set of conclusions, as a means to stimulate further debate on this topical issue.
Radiation Effects on Materials in the Near-Field of a Nuclear Waste Repository
Authors: Gu B.X. , Wang L.M. , Wang S.X. , Ewing R.C.
Published: May 2011

The long-term radiation effects on materials in the near-field of a nuclear waste repository have been evaluated using accelerated laboratory experiments with energetic electron or ion beam irradiation. The materials studied include: zeolites, layered silicates (smectite clay and mica), as well as crystalline silicotitanate (CST) which is an important ion exchange material for the chemical separation of high-level liquid radioactive wastes.

In situ transmission electron microscopy (TEM) during irradiation by energetic electrons and ions has shown that all of the studied materials are susceptible to irradiation-induced amorphization. At room temperature, complete amorphization was observed after ionizing doses of 1010 ∼ 1012 Gy or displacement doses on the order of 0.1 dpa (equivalent to doses received in 400-1,000 years for a high-loading nuclear waste form). Amorphization may be preceded or accompanied by dehydration, layer spacing reduction and gas bubble formation. In the case of zeolites, CST and some layered silicates, radiation effects are significantly enhanced at higher temperatures. Our experiments have shown that amorphization or even partial amorphization will cause a dramatic reduction in ion exchange and sorption/desorption capacities for radionuclides, such as Cs and Sr. Because the near-field or chemical processing materials (e.g. zeolites or CST) will receive a substantial radiation dose after they have incorporated radionuclides, our results suggest that radiation effects may, in some cases, retard the release of sorbed or ion-exchanged radionuclides.
Partitioning of Actinides, Rare Earth Elements, and Other Trace Elements In Titanium-Rich Veins From Adamello, Italy
Authors: Lumpkin Gregory R. , Gieré Reto , Payne T.E. , McGlinn P.J. , Hart K.P.
Published: May 2011

Extensive mineralogical and chemical studies have been carried out on the Ti-rich hydrothermal veins emplaced within the contact aureole of the Adamello batholith. In addition to other actinide and rare earth element host phases, the veins contain both zirconolite and betafite and provide information relevant to ceramic wasteforms designed for the disposal of actinide-rich nuclear wastes. In this paper, we describe the results of element partitioning studies based on dissolution experiments using 9M HCl. Generally, the acid-resistant minerals include allanite, baddeleyite, betafite, chalcopyrite, geikielite, titanite, spinel, and zirconolite. We also found that the major silicate minerals forsterite, phlogopite, and titanian clinohumite and the sulfide mineral pyrrhotite are partially dissolved by the acid treatment, whereas calcite and apatite are highly soluble (as expected). In particular, the distributions of Th and U between the acid-resistant and acid-soluble fractions indicate that they partition mainly between zirconolite, titanite, betafite, and apatite. However, there is a considerable increase in the amounts of Zr, Nb, Th, and U released in certain actinide-rich samples that may result from enhanced dissolution following radiation damage.
Surveying for Migration Pathways in the Granitic Rock Using Nuclear Track Detectors, Autoradiography and Digital Image Analysis as an Aid To Construct the Basis for Heterogeneous Diffusion Modeling
Authors: Kemppainen M. , Oila E. , Siitari-Kauppi M. , Sardini P. , Hellmuth K-H.
Published: May 2011

Radioelement migration within a rock matrix under natural long-term conditions is a complex process controlled by various parameters. Pure physical parameters such as porosity, hydraulic conductivity and diffusivity are usually sufficient to describe transport in well-defined laboratory systems. In natural rock matrices transport is influenced by physical pore properties such as pore size distribution, connectivity, tortuosity, constrictivity and petrological and chemical nature and charge on the fluid-rock interface. The overall characterization of heterogeneous rock structures is needed for the accurate heterogeneous diffusion modeling.

Here we describe a method for the detection of μ-particles from uranium in cm-scale rock samples based on the analysis of the tracks formed in organic polymer, CR-39. On the other hand the uranium tracks were compared with the migration pathways and porosity distribution produced with the 14C-polymethylmethacrylate impregnation method (14C-PMMA). For analyzing mineral specific uranium occurrence and porosities the staining methods were used to produce the mineral map of the rock sample. Digital image analysis techniques were applied to the different cm-scale pictures of rock samples. Scanning electron microscopy and energy dispersive X-ray analysis (SEM/EDS) were performed in order both to study the pore apertures of grain boundaries and fissures in greater detail and to detect the uranium phases.

The high uranium content was found to be congruent with the porous mineral phases; altered plagioclase and biotite grains, and the intra- and intergranular fissures detected with the 14C-PMMA technique. Plenty of microfractures transsecting potassium feldspar and quartz grains were filled with calcite together with precipitated uranium.
Non-Partisan Candidates and Lists at Slovenian Local Elections, 1994-2010
Authors: Kukovic Simona , Hacek Miro
Published: May 2011

The paper focuses on a lesser-known political phenomenon observed in Slovenia since the country gained its independence in 1991. At every local election since then, non-partisan candidates and lists—often called independent in the media—have been gaining more votes and increasing support. By analyzing the results of the last five local elections, we manifested that there are three origins of the success of these non-partisan candidates and lists. We also try to ascertain whether non-partisan candidates and lists are truly a product of an anti-party political culture, climate and movements, or whether they are simply another way for political parties to gain political power at local levels of government.