bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Review of Corrosion Considerations of Container Materials Relevant to Underground Disposal of High-Level Radioactive Waste in Belgium
Authors: Kursten Bruno , Druyts Frank 
Published: April 2011

ABSTRACTThe underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.
Interaction Swelling Bitumen – Host Formation: Scoping Calculations
Authors: Li Xiang Ling , Bernier Frédéric , Valcke Elie 
Published: April 2011

ABSTRACTThe Eurobitum bituminised waste product (BWP) from the former EUROCHEMIC reprocessing plant in Mol-Dessel (Belgium) contains a lot of salts, mainly NaNO3 and CaSO4. In contact with pore water in an underground repository, the dehydrated salts in the BWP will rehydrate, resulting in swelling and, possibly, a swelling pressure build-up. A high swelling pressure might impede the integrity and the safety of the repository by creation of preferential pathways for radionuclide migration. In Belgium, the Boom Clay is studied as a reference host formation.The interaction between the swelling BWP and the host formation is a very complicated hydromechanical process. It depends not only on the hydromechanical behaviour of the Boom Clay, but also on that of the BWP. The hydromechanical constitutive law of Eurobitum is not yet established. We have therefore realised scoping calculations to assess the stress redistribution and deformation due to the recompression of the clay around a disposal gallery caused by a swelling pressure. Sensitivity studies have evidenced some important influencing factors.The scoping calculations showed that the swelling pressure exercised to the Boom Clay should be limited to 7 - 8 MPa. Beyond this limit, the Boom Clay host formation will be subject to important deformation, and the risk of preferential paths for radionuclide transport becomes very high. The convergence during the excavation of the disposal galleries is of primary importance for a safe waste disposal system design. Indeed, excess convergence may result in a higher damaged zone before disposal of the bituminised waste.
Bituminized Waste Leaching in Restricted and Free Swelling Conditions: Mechanisms and Analytical Modeling
Authors: Lefebvre X. , Sercombe J. , Ledieu A. , Gwinner B. , Adenot F. 
Published: April 2011

ABSTRACTThe leaching behavior of nuclear wastes embedded in bitumen has been extensively studied for a decade by many teams of researchers. The main conclusion is that both water uptake and salts release obey to a diffusive regime. The most preponderant parameters of the Bituminized Waste Products (BWPs) leaching are, on the one hand, the nature and the content of the salts that are embedded in the bitumen, and on the other hand, the microstructure of bitumen and the external constraint applied to this confining material.The numerical code COLONBO, designed by the CEA, allows to predict the behavior of the BWPs leaching in the case of salts embedded in soft bitumen and in free swelling conditions (no mechanical constraints during the leaching). This paper intends to achieve two goals: (i) to propose an analytical resolution of the system of equations implemented in COLONBO by using accurate simplification. The validation of this approach relies on the comparison with the predicted solutions of COLONBO and, when available, with the experimental results. (ii) To extend this simplified modeling to the case of restricted swelling conditions. The point is that the validation of this approach would allow us to propose a unified system of equations, allowing to simulate the BWPs leaching in whatever conditions and improving consequently the field of validity of the predictions for the behavior of bituminized waste products.
Mechanisms and Modelling of Waste/Cement Interactions – Survey of Topics Presented at the Meiringen Workshop
Authors: Wieland E. , Johnson C.A. , Lothenbach B. , Winnefeld F. 
Published: April 2011

ABSTRACTCementitious matrices are being used worldwide as a containment medium for radioactive and non-radioactive waste in order to retard the mobility of contaminants. The present thrust of research is to further the understanding of contaminant binding in the cementitious matrix in order to predict the long-term behaviour and the potential impact of the waste on the environment.The workshop “Mechanisms and Modelling of Waste/Cement Interactions”, held in Meiringen, Switzerland, between May 8 and 12, 2005, focused on the chemical understanding and thermodynamic modelling of the processes responsible for the retention of radioactive and non-radioactive species in cementitious systems. The objectives of the workshop were to bring together scientists from different disciplines, i.e. cement chemistry, radioactive and non-radioactive hazardous waste disposal, to stimulate discussions on current developments and to identify future needs in this field of research. The topics treated in the workshop were chosen to maximize the benefit to the different fields of research. Cement chemists reported on developments in the understanding of cement mineralogy and thermodynamic modelling of cement systems. The hazardous and radioactive waste management communities presented their ideas on the mechanisms of contaminant binding to cement minerals as well as field, laboratory and modelling results from practical applications. In this paper important areas of research on waste/cement interactions presented in the workshop will be outlined and briefly discussed. The following overview reflects a subjective perception of the workshop and does not lay claim to deal comprehensively with all the papers that were presented in the workshop.
Molten ceramic solidification during molten state processing of HLW
Authors: Kinoshita Hajime , Uno Masayoshi , Yamanaka Shinsuke , Lee William E. 
Published: April 2011

ABSTRACTSolidification of molten ceramics during molten state processing of High-Level Waste (HLW) was studied to investigate a possibility of a simpler and flexible process for immobilisation of HLW. Simulated HLW was heated together with different reducing agents (TiN with AlN or TiO2) either at 1673 K or 1873K for 1 to 3 hours under Ar atmosphere. Cooling rates of 5 K/min and 10 K/min did not result in any obvious differences in the products obtained under the standard condition used in the present study. The composition and quantity of the reducing agent had a significant effect on the melting and solidifying behaviour of the simulated HLW. The results suggest the possible further studies to improve the solidification process and the final product.
Mechanical integrity of yttria-stabilised zirconia doped with Np oxide
Authors: Kinoshita Hajime , Kuramoto Ken-ichi , Uno Masayoshi , Yamanaka Shinsuke , Mitamura Hisayoshi , Banba Tsunetaka 
Published: April 2011

ABSTRACTMechanical properties of YSZ doped with Np oxide were studied to investigate the sufficiency to be a waste form for immobilisation of highly concentrated TRU. The study was conducted focusing on Vickers (HV) and Knoop (HK) hardness, Young's modulus (E) and fracture toughness (KIC). The results showed that YSZ is harder and more resistant to elastic deformation and crack development than such waste forms as borosilicate glass and synroc. The effects of porosity and Np content on HV, HK, E and KIC are also discussed.
Investigation of Natural and Artificial Zr-silicate Gels
Authors: Burakov Boris E. , Smetannikov Andrey Ph. , Anderson Evgeniy B. , Alexeev Alexander Yu. 
Published: April 2011

ABSTRACTSample of natural Zr-silicate gel containing up to 13 wt.% U was characterized using scanning electron microscopy (SEM), X-ray powder diffraction (XRD), high resolution transmission electron microscopy (HRTEM) and electron microprobe analysis (EPMA) method. It was found that gel matrix is amorphous in general; however, it contains non-identified nanocrystallites. No separated oxide phases of U, Zr or Si were observed in a gel matrix. After sintering in air at 1400°C for 1 hour gel transformed largely into crystalline zircon, (Zr,U)SiO4. Uranium was not found in any other phases besides zircon. It was assumed that high chemical durability of natural Zr-P-U-Ti-silicate gel is caused by two competing processes which exist under self-irradiation conditions: 1) crystallization of the gel and 2) metamictization of the crystallized zircon and other phases.Synthetic samples of Zr-silicate gel doped with Ce, U, Pu and Am were obtained and studied in comparison with natural samples. It was suggested to use artificial solid Zr-silicate gels for durable fixation of actinides for the goal of long-term or intermediate storage.
Leaching behaviour of non-irradiated and irradiated HTR UO2-ThO2fuel particles under reducing conditions
Authors: Alliot Cyrille , Grambow Bernd 
Published: April 2011

ABSTRACTThe dissolution of irradiated HTR UO2-ThO2fuel particle under reducing conditions was studied using a continuous flow-through reactor and compared to the dissolution of unirradiated fuel particle in the same condition. The irradiated fuel particle was leached for more than eight months. A fast137Cs release was observed, corresponding to a labile fraction (“instant release fraction”). Then, a congruent leaching of137Cs and90Sr was measured corresponding to a matrix dissolution rate equal to 1.7 mg/m2/d. A slower release of238Pu probably due to sorption phenomena is observed. Dissolution rates are 100-2000 times higher than for unirradiated material. We can conclude that alpha-radiolysis is responsible for this increase due to local oxic conditions. However we cannot exclude that due to irradiation the accessible surface area has been increased as well.The coating of this irradiated particle was equally studied to determine the presence of different radionuclides and their leaching properties of leaching
Oxidation kinetics of high burn-up UOX and MOX fuel : the influence of the grain boundaries
Authors: Desgranges Lionel , Rousseau Gurvan , Ferroud-Plattet Marie-Pierre , Ferry Cécile , Giaccalone Hélène , Aubrun Isabelle , Delion Pierre , Untrau Jean-Michel 
Published: April 2011

ABSTRACTIn a dry air storage facility, the oxidation of irradiated nuclear fuel is considered as the main risk induced by a leakage that would put the nuclear ceramics in contact with air. The oxidation of UO2 in U3O8 leads indeed to a relative volume swelling of more than 30% which could cause the ruin of the fuel rod. In order to fill the lack of data on high burn-up fuel, some oxidation experiments are currently conducted in CEA with UO2 and MOX fuel. Fragments as well as slices of fuel rod, with a high burn-up, have been oxidized in air at 200°C for more than 4000 hours. The associated weight gain curves are presented together with the characterization of some samples picked up at different oxidation times. It is shown that the weight gain curve of high burn-up UOX is not parabolic-shaped, as predicted in literature, but sigmoid-shaped. Moreover it is also shown that the grain boundary oxidation is not uniform inside a fragment. This leads to a new description of the oxidation mechanism of irradiated fuel.
Testing of uranium dioxide enriched with 233U under reducing conditions
Authors: Ollila K. , Oversby V. 
Published: April 2011

ABSTRACTExperiments have been conducted to determine an upper limit to the dissolution rate of UO2 under reducing conditions at great depths in granitic bedrock. The aqueous phase was a dilute, synthetic groundwater that simulates fresh groundwater conditions, so-called modified Allard groundwater. Four dissolution tests of long duration were performed. Test duration ranged from 52 to 140 days. Reducing conditions were produced with metallic iron. Average dissolution rates for the UO2 samples under the test conditions were preliminarily calculated based on the total recovery of U at the end of the tests. The dissolution rates ranged from 0.12 to 0.14 ppm/yr for the sum of the total release from the third and fourth cycles of testing (269 days). There was no indication of an effect of alpha radiolysis on the dissolution rate results for UO2 samples with 0, 5, and 10% 233U.
Study of simulated HLW-glass leaching behavior under low oxygen repository condition
Authors: Hua Zhang , Shanggeng Luo 
Published: April 2011

ABSTRACTMany countries have recognized vitrification as the suitable solidification method for conditioning high level radioactive waste before they are isolated from biosphere in geologic disposal. This paper studies the leaching behavior of a simulated HLW-glass (90Nd/10) under low oxygen repository condition during 1.5 years. The leaching results showed that the glass was corroded less under low oxygen condition than under ambient atmosphere. XRD analyses showed some little diffraction peak appeared on the leached glass surface at 150°C. At the same time, no diffraction peak appeared on the leached glass surface at 90°C. The secondary products formed on the surface of the leached glass contained mainly Si, Al, Ca and Fe.
Kinetics of Aqueous Alteration of P0798 Simulated Waste Glass in the Presence of Bentonite
Authors: Yamaguchi K. , Inagaki Y. , Saruwatari T. , Idemitsu K. , Arima T. , Yoshikawa H. , Yui M. 
Published: April 2011

ABSTRACTStatic aqueous alteration tests were performed with a Japanese simulated HLW glass, P0798, in the presence of bentonite in order to understand the effects of bentonite on the glass alteration kinetics and on the associated Cs release. Analogous alteration tests were performed in 0.001M NaOH solution without bentonite for comparison. The results indicated that; 1) at the initial stage of alteration up to 50 days, no remarkable difference was observed in the alteration rate between both cases “with” and “without” bentonite, 2) at the later stage beyond 50 days, however, the rate in the case “with” bentonite was larger than that in the case “without” bentonite. These results on the alteration rate were analyzed by use of a water-diffusion model. In the case “without” bentonite, a good agreement was observed between the model analysis and the experimental results at the initial stage of alteration up to 50 days, however, the model analysis deviated from the experimental results at the later stage beyond 50 days. In the case “with” bentonite, on the other hand, a good agreement was observed even at the later stage to give the value of the apparent diffusion coefficient, Di of 3.5×10−21m2/s. The comparison between both cases suggests that the alteration rate is controlled by the water diffusion in both cases “with” and “without” bentonite, however, the rate is depressed in the case “without” bentonite probably by the protective effects of the alteration layer developing at the glass surface. In the case “with” bentonite, on the other hand, the alteration layer is expected to be less protective. Cesium desorption tests performed for the altered glass and bentonite indicated that most of the cesium dissolved from the glass is retained in the secondary phase of smectite developing in the precipitated layer by sorption with ion-exchange in the case “without” bentonite. In the case “with” bentonite, however, it is likely to be sorbed at bentonite surface.
Assessing the durability of nuclear glass with respect to silica controlling processes in a clayey underground disposal
Authors: Windt Laurent De , Leclercq Stéphanie , van der Lee Jan 
Published: April 2011

ABSTRACTThe long-term behaviour of vitrified high-level waste in an underground clay repository was assessed by using the reactive transport model HYTEC with respect to silica diffusion, sorption and precipitation processes. Special attention was given to the chemical interactions between glass, corroded steel and the host-rock considering realistic time scale and repository design. A kinetic and congruent dissolution law of R7T7 nuclear glass was used assuming a first-order dissolution rate, which is chemistry dependent, as well as a long-term residual rate. Without silica sorption and precipitation, glass dissolution is diffusion-driven and the fraction of altered glass after 100,000 years ranges from 5% to 50% depending on the fracturation degree of the glass block. Corrosion products may limit glass dissolution by controlling silica diffusion, whereas silica sorption on such products has almost no effect on glass durability. Within the clayey host-rock, precipitation of silicate minerals such as chalcedony may affect glass durability much more significantly than sorption. In that case, however, a concomitant porosity drop is predicted that could progressively reduce silica diffusion and subsequent glass alteration.
Chemical durability of high-level waste glass in repository environment: main conclusions and remaining uncertainties from the GLASTAB and GLAMOR projects
Authors: Iseghem P. Van , Lemmens K. , Aertsens M. , Gin S. , Ribet I. , Grambow B. , Crovisier J.L. , Nero M. Del , Curti E. , Schwyn B. , Luckscheiter B. , McMenamin T. 
Published: April 2011

ABSTRACTThis paper reviews the main conclusions of two European Commission funded projects, GLASTAB and GLAMOR, on the durability of HLW glass in geological disposal, and the remaining uncertainties. The progress in GLASTAB relates to the characterization of the alteration layer, the modeling of glass dissolution, the interaction of glass with near field materials, the behaviour of radionuclides, and the calculation of HLW glass dissolution as part of geological disposal. The GLAMOR programme was focused on the decrease of the glass dissolution rate upon achieving silica saturation in solution. Two conclusions from this programme are the importance of the residual rate measured beyond silica saturation and the need to reduce the uncertainties in model parameters. The overall conclusion from the two projects is that strong progress has been achieved on basic dissolution mechanisms and interaction with disposal environments - HLW glass will act as a good barrier to the release of radionuclides in geologic disposal.
Development and application of knowledge-based source-term models for radionuclide mobilisation from contaminated concrete
Authors: Deissmann Guido , Bath Adrian , Jefferis Stephan , Thierfeldt Stefan , Wörlen Stefan 
Published: April 2011

ABSTRACTConcrete materials in nuclear facilities may become activated or contaminated by various radionuclides through different mechanisms. Consequently, decommissioning and dismantling of these facilities produce considerable quantities of these materials (e.g. concrete structures, rubble), which are at least potentially contaminated with radionuclides and which must be managed safely and cost-effectively. In this paper, we present results from a research project that aims at the development of source-term models for the mobilization of radionuclides from contaminated concrete. The objective of this task was to clarify whether a more realistic source-term description could be beneficial for optimization of the management of decommissioning wastes by reducing the amount of material for disposal as radioactive waste as well as by saving natural resources due to the recycling of building materials.To identify important parameters and processes that affect the release rates of radionuclides, we evaluated the chemical behavior and the solid speciation of radionuclides in concrete materials and the influence of factors like concrete properties, source/pathway of contamination, and the scenario-specific chemical environment and hydraulic regime. Furthermore, concrete degradation processes and their influence on contaminant mobilization were addressed. On this basis, source-term models were developed to describe the radionuclide release by (i) the dissolution of radionuclide containing solid phases, (ii) the desorption of radionuclides from surfaces, and/or (iii) the leaching of radionuclides from a solid matrix without disrupting its structure. These source-term models were parameterized for probabilistic simulations of various release options, including the reuse of recycled building materials, the disposal of rubble in inert and municipal landfills as well as the on-site disposal of concrete materials (e.g. foundations remaining in the ground, in situ burial of rubble). For some scenarios and radionuclides, the calculated release rates were between one and two orders of magnitude lower than those used in former generic calculations. Based on the results of stochastic simulations, the consequences of the use of a more realistic source-term for dose assessments with respect to clearance/recycling of contaminated concrete will be illustrated and discussed.
Magnetic anisotropy, sress, and martensitic transformation in Ni-Mn-Ga thin films on Si(100) wafer
Authors: Hagler Michael , Chernenko Volodymyr A. , Ohtsuka Makoto , Besseghini Stefano , M¨llner Peter 
Published: April 2011

ABSTRACTNi-Mn-Ga magnetic shape memory alloys (MSMAs) tend to undergo a large deformation upon the application of a magnetic field. This deformation is attributed to twin boundary motion in the martensitic phase. In an effort to harness the shape memory effect for use in sensors, actuators, and micro-devices, the behavior of Ni-Mn-Ga thin films is attracting attention. Substrate curvature measurements were done with Ni-Mn-Ga films with a thickness of 2.0 μm sputter-deposited on Si(100) wafer having amorphous 500 nm thick SiNx buffer layer. During the wafer bow curvature measurements, stress levels of 0.65 GPa were attained. The martensitic transformation is manifested by a stress-temperature hysteretic loop. Measurements of magnetization curves were carried out on Ni-Mn-Ga films with thickness between 0.5 and 3.0 μm. A change of the magnetization behavior from the easy-plane type for thin films to the out-of-plane easy-axis type for thick films is observed. This effect is caused by the interplay between different contributions to the overall anisotropy of film.