bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

ABSTRACTSeveral 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

ABSTRACTExperimental 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

ABSTRACTOne 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

ABSTRACTZirconolite-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

ABSTRACTAt 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

ABSTRACTThis 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

ABSTRACTThe 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

ABSTRACTUO2 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

ABSTRACTLeaching 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

ABSTRACTAlloy 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

ABSTRACTIn 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTDiffusion 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

ABSTRACTThe 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

ABSTRACTIn 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.