bluuebunny/crossref_metadata_2025_split huggingface dataset:
Building Confidence in Radionuclide Transport Models for Fractured Rock: The Nagra/JNC Radionuclide Retardation Programme
Authors: Ota K. , Alexander W.R. , Smith P.A. , Möri A. , Frieg B. , Frick U. , Umeki H. , Amano K. , Cowper M.M. , Berry J.A. 
Published: May 2011

ABSTRACTThe joint Nagra/JNC Radionuclide Retardation Programme has now been ongoing for 15 years with the main aim of direct testing of radionuclide transport models in as realistic a manner as possible. A large programme of field, laboratory and natural analogue studies has been carried out at the Grimsel Test Site in the central Swiss Alps and the Kamaishi In Situ Test Site in north-east Japan. The understanding and modelling of both the processes and the structures influencing radionuclide transport/retardation in fractured host rocks have matured as has the experimental technology, which has contributed to develop confidence in the applicability of the underlying research models in a repository performance assessment. In this paper, the successes and set-backs of this programme are discussed as is the general approach to the thorough testing of the process models and of model assumptions. In addition, a set of key findings is presented, involving discussions on the enhancement of confidence through the program.
On the Hydrolysis of Tetravalent Metal Ions
Authors: Ekberg C. , Brown P. , Comarmond J. , Albinsson Y. 
Published: May 2011

ABSTRACTThe stability constants of the hydroxide complexes of zirconium(IV) and uranium(IV) have been measured at 15, 25 and 35°C [in 1.0 mol dm−3 (Na, H)ClO4] using potentiometry. For zirconium(IV), the results indicate the presence of Zr(OH)3+ and the polymeric species Zr3(OH)48+ and Zr4(OH)88+ whereas the results for uranium(IV) indicate the presence of U(OH)3+ and the polymeric species U4(OH)124+. The hydrolysis of both metal ions was studied at three temperatures allowing the determination of ΔH° and ΔS° of reaction for each species. The results were compared with previous results determined for thorium(IV) under the same experimental conditions to ascertain whether thorium should be used as an analogue for other tetravalent metal ions in safety assessment studies of nuclear repositories.
Categorization of Cement Hydrates by Radionuclide Sorption Mechanism
Authors: Noshita K. , Nishi T. , Yoshida T. , Fujihara H. , Saito N. , Tanaka S. 
Published: May 2011

ABSTRACTA categorization of cement components was attempted from the viewpoints of their chemical constituents and sorption behaviors. In this study, we assumed that the cement components are categorized into four groups, calcium silicate compounds, hydroxides, calcium aluminate compounds, and others. The sorption behaviors were compared among the four groups by batch sorption experiments carried out for 12 kinds of cement components using the radionuclides cesium, iodine, inorganic carbon, and nickel as different chemical species. The characteristic sorption trends were obtained in each group. We found that the sorption mechanism is closely related to the chemical constituents and surface species. Categorization of these cement of components will contribute to the understanding of sorption mechanisms.
Sorption Behavior of Radium and Actinium on Soils
Authors: Sakamoto Yoshiaki , Takebe Shinichi , Ogawa Hiromichi , Muraoka Susumu , Ishii Tomoaki , Inagawa Satoshi , Gunji Yasuyoshi 
Published: May 2011

ABSTRACTDistribution coefficients of 226Ra and 227Ac on soils and tuff have been obtained by a batch method. The distribution coefficients of Ra on loam, two kinds of sand and tuff were 28m3/kg, 0.45m3/kg, 0.19m3/kg and 1.9m3/kg, respectively. A linear relationship between the distribution coefficients of Ra on each soil and Cation Exchange Capacity(CEC) of each soil was obtained to suggest cation exchange sorption reaction of Ra2+ on soil surface. Sorption-migration behavior of Ra in sand column has been studied. Most of Ra remained in the column after passing of 5,000 pore volume of the column, showing several times the distribution coefficients of Ra than that by the batch method. The distribution coefficients of 227Ac on the soil samples were in the range from 1m3/kg to 10m3/kg.
Diffusion and Adsorption in Porous Silica Considered as a Reference Material for Crystalline Rock
Authors: Aalto Hannu , Lehikoinen Jarmo , Muurinen Arto , Valkiainen Matti 
Published: May 2011

ABSTRACTIn the present study, well-characterized mesoporous silica was used as a reference material for crystalline rock. Saturation-leaching and through-diffusion tracer experiments were systematically carried out to determine the adsorption capacity factors and apparent and effective diffusivities for the radioactive tracers 3H, 22Na and 36Cl. The ionic strength of the supporting NaCl electrolyte, in equilibrium with atmospheric air, was either 0.002 M or 0.1 M. The apparent diffusivities derived from saturation-leaching experiments reveal that equilibrium adsorption on a negatively charged silica surface does not act to retard the mass transfer of sodium ions. The steady-state diffusion results are found to demonstrate the increasing break-through for sodium with decreasing ionic strength of the background electrolyte in comparison to 3H. These observations are consistent with the picture of chloride ion exclusion and of the diffuse double layer surface-excess for sodium ions being transported in the direction of the macroscopic concentration gradient. Furthermore, they are shown to be at odds with the widely adopted macroscopic pore diffusion model, which neglects the mobility of the counter-ion surface- excess.
Long-Term Behavior of Bituminized Waste: Modelling of Auto-Irradiation and Leaching
Authors: Chaix P. , Camaro S. , Simondi B. , Vistoli P.P. , Blanc V. 
Published: May 2011

ABSTRACTDuring interim storage, the main evolution factor of the bituminized wastes is radiolysis under auto-irradiation, while in deep disposal conditions, after radioactive decrease, site resaturation with water, and container corrosion, the main factor of evolution is leaching by water.This paper presents two models, designed to describe respectively radiolytic gas evacuation, and bituminized waste alteration under leaching.The first model is based on the integration of simple descriptions of gas production by radiolysis, gas diffusive transport, bubble nucleation and ripening, and bubble transport under buoyancy and viscosity.The second model is based on simple descriptions of water adsorption and solubilization at the surface, water diffusion into the matrix, water uptake by the most soluble salts, salt grain swelling and alteration of matrix waterproofing, diffusive transport of salts out of the waste.
The Effect of Radiolytic Degradation Products of Eurobitum on the Solubility and Sorption of Pu and Am in Boom Clay
Authors: Valcke E. , Sneyers A. , Iseghem P. Van 
Published: May 2011

ABSTRACTThe potential effect of water-soluble organic ligands, originating from the radiolytic degradation of Eurobitum bituminized waste, on the Pu and Am solubility and sorption in Boom Clay, was assessed. Granulated inactive Eurobitum simulate in contact with representative solutions was gamma-irradiated, and the irradiated solutions were thoroughly analyzed (GC-MS, IC, ICP-AES,…). Pu and Am solubility and sorption tests were then performed in pure Real Boom Clay Water (RCW) and in RCW containing 50 mass% of the solution containing the radiolytic degradation products. The relevance of the applied methodology is discussed.
Neodymium Incorporation in Zirconolite-Based Glass-Ceramics
Authors: Loiseau P. , Caurant D. , Baffier N. , Fillet C. 
Published: May 2011

ABSTRACTThe investigations on enhanced reprocessing of nuclear spent fuel, and notably on separating the long-lived minor actinides, such as Am and Cm, from the other fission products have led to the development of highly durable specific matrices such as glass-ceramics for their immobilization. This study deals with the characterization of zirconolite (CaZrTi2O7) based glass-ceramics synthesized by devitrification of an aluminosilicate parent glass. Trivalent actinide ions were simulated by neodymium, which is a paramagnetic local probe. Glass-ceramics with Nd2O3 contents ranging from 0 to 10 weight % were prepared by heat treatment of a parent glass at two different growth temperatures: 1050° and 1200°C. X-ray diffraction (XRD), energy dispersive X-ray analysis (EDX) and electron spin resonance (ESR) measurements clearly indicate that Nd3+ ions are partly incorporated in zirconolite crystals formed in the bulk of the glass-ceramic samples. The amount of neodymium in the crystalline phase was estimated using ESR results and was found to decrease with increasing either heat treatment temperature or total Nd2O3 content.
Thermochemical Approach on Treating Spent Ion Exchange Resins
Authors: Ojovan M.I. , Klimov V.l. , Petrov G.A. , Dmitriev S.A. , Laurson A.V. , Semenov K.N. 
Published: May 2011

ABSTRACTA thermochemical approach was suggested for treating and conditioning specific streams of radioactive wastes for example spent ion exchange resins (IER), mixed, organic or chlorine-containing radioactive waste (PVC, etc.). Conventional thermal treatment of such waste encounters serious problems concerning complete destruction of organic molecules and possible emissions of radionuclides, heavy metals and chemically hazardous species. The thermochemical treatment uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. The composition of the PMF is designed in such a way as to minimize the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the ash residue. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. Previous preliminary work demonstrated the feasibility of applying the thermochemical approach to treatment of spent IER using PMF. Herein, the results are presented of theoretical and experimental studies to define the optimal PMF composition as well as the PMF/waste ratio.
The Effect of Gamma Radiation on the Dissolution of High-Level Waste Glass in Boom Clay
Authors: Lemmens K. , Iseghem P. Van 
Published: May 2011

ABSTRACTThe effect of γ-radiation on the dissolution of candidate high-level waste glasses was investigated in potential disposal environments at 90°C. The media consisted of mixtures of Boom Clay, bentonite clay and cement with clay water. During the experiments the pH of Boom Clay decreased, probably mainly by radiolytical oxidation of pyrite. The addition of bentonite, cement and glass buffered the pH decrease. Under radiation the glass mass losses decreased, whereas the leach rate of soluble elements was not influenced or appeared to increase. This is explained through the radiolytical acidification, and possibly by bubble formation in the glass.
Surface Alteration of Nd-Bearing Zirconolite Following Hydrothermal Treatment
Authors: Zhang Zhaoming , Vance Eric R. , Blackford Mark G. 
Published: May 2011

ABSTRACTStatic dissolution tests on Nd-bearing zirconolite were conducted in deionized water at 150°C for up to 6 months. Surfaces, both before and after aqueous dissolutions, were examined using X-ray Photoelectron Spectroscopy (XPS) and Scanning Electron Microscopy (SEM). Individual submicron-sized crystals wereobserved only on some zirconolite grains after hydrothermal treatment for one week, and were identified as brookite (possibly plus anatase) by Transmission Electron Microscopy (TEM). The number of secondary crystals present on the zirconolite surface was, however, reduced significantly after 6 months of durability testing. The results of electron microscopy studies are consistent with those obtained from XPS in that the average Ca and Ti concentrations of the top surface layer (< 5 nm) decreased slightly with respect to Zr, while the average Al concentration increased after durability testing.
Feasibility Limits in Using Cerium as a Surrogate for Plutonium Incorporation in Zircon, Zirconia and Pyrochlore
Authors: Zamoryanskaya M.V. , Burakov B.E. 
Published: May 2011

ABSTRACTCerium is a well-known surrogate element for Pu that is widely used in experiments on the synthesis and study of ceramic waste forms for actinide immobilization. Cathodoluminescence (CL) images and emission spectra from synthetic zircon, (Zr,…)SiO4, cubic zirconia, (Zr,Y,Gd,…)O2 and pyrochlore, Ca(Hf,Gd,…)Ti2O7, individually doped with Ce and Pu were investigated. It was determined that some of the Ce, which was added as Ce4+ in the form of CeO2, was converted to the 3+ valence state during synthesis of zircon-based ceramics by sintering in air and in argon gas. However, Pu-doped zircon synthesized in air under the same conditions does not exhibit the CL emission of Pu3+. Analysis of CL spectra of cubic zirconia doped with Ce as well as Pu suggests that both ions are in the tetra-valent state in these synthetic materials. However, the CL peak of Ce-doped zirconia is almost identical to that of pure zirconia (2.6 eV). The CL peak for Pu-doped zirconia is shifted to 2.4 eV. No CL emission of Ce3+ or Pu3+ was observed in pyrochlore samples. The CL spectra for Ce-and Pu-doped pyrochlore samples were almost identical. We suggest that under oxidizing conditions Ce is good surrogate of Pu for experiments on the synthesis of pyrochlore-based and acceptable enough for zirconia- based ceramics. The use of Ce as Pu surrogate for zircon-based ceramics should be considered as limited.
Effect of Synthesis Conditions on Phase Composition of Pyrochlore-Brannerite Ceramics
Authors: Stefanovsky S.V. , Yudintsev S.V. , Nikonov B.S. , Lapina M.I. , Aloy A.S. 
Published: May 2011

ABSTRACTThree melted samples of pyrochlore-brannerite-based ceramics produced under different redox conditions were examined. Two of the samples were produced using cold crucible melting at ~1600 C. The third sample was obtained via melting in a microwave oven at 1700-1800 C. All the samples are composed of major pyrochlore and brannerite phases, and minor rutile and UO2-based solid solution or pseudobrookite phases. Pyrochlore-structured phases predominate in all three samples and account for 50-60% of the total bulk. Two pyrochlore varieties – Ca-pyrochlore (predominant) and Ba-pyrochlore have been found in these samples. The latter phase is more stable at high temperatures than the Ba-hollandite present in sintered pyrochlore-rich Synroc-F ceramics. Decomposition of the Ba-hollandite results in rutile formation in the melted samples.
Pyrochlore-Rich Synroc as a Host for Immobilization of Actinides
Authors: Yang Jianwen , Tang Baolong , Luo Shanggeng 
Published: May 2011

ABSTRACTSix different formulations of pyrochlore-rich synroc waste forms were designed to contain 32-50wt% simulated actinides. Their physical properties were determined. Phase identification was carried out by using X-ray diffractometer (XRD), and backscattered electron image (BEI) in scanning electron microscopy (SEM). Phase microanalyses were carried out using energy dispersive spectrometer (EDS) in SEM. Their durability properties were determined by PCT leach tests, the leachate was analyzed by using inductive coupled plasma mass spectroscopy (ICP/MS) and atomic absorption spectroscopy (AAS).
Aqueous Dissolution of Rb-Bearing Hollandite and Synroc-C at 90°C
Authors: Carter M. L. , Vance E. R. , Lumpkin G. R. , Loi E. 
Published: May 2011

ABSTRACTRb-doped Ba-hollandite and synroc-C samples were prepared for aqueous dissolution studies at 90°C in deionized water using MCC-1 protocols. Small amounts of possibly Rb-bearing BaTiO3, hibonite and alumina accompanied the major Rb-doped (0.1 formula units) hollandite phase. The Rb loadings in the synroc samples ranged from 0.89 to 2.67 wt%. Rb leached at <0.2 g/m2/d from hollandite samples and <0.1g/m2d from synroc, with the values decreasing with increasing dissolution time. The variance between the dissolution rates for the different Rb-doped hollandite and synroc samples is discussed, in terms of Rb-bearing minor phases.
Effect of Spent Fuel Burnup and Composition on Alteration of the U(Pu)O2 Matrix
Authors: Jégou C. , Peuget S. , Lucchini J.F. , Corbel C. , Broudic V. , Bart J.M. 
Published: May 2011

ABSTRACTFor a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified.Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd·tHM−1 and a MOX fuel sample irradiated to 47 GWd·tHM−1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25°C) in a hot cell.After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 à 10−7 d−1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings.
Long-Term Dissolution Behavior of Spent Fuel in Compacted Bentonite and Synthetic Granitic Groundwater
Authors: Chun K.S. , Kim S.S. , Kang C.H. 
Published: May 2011

ABSTRACTThe long-term dissolution behavior of spent PWR fuel in synthetic granitic groundwater has been investigated since June of 1998 in order to identify the release mechanism of spent fuel in contact with domestic Ca-bentonite, which was compacted as a density of 1.4g/cm3. Several spent fuel specimens were cut to around 3-mm thick, and these specimens were then loaded into each leaching cell. Every half year, one specimen was collected and the bentonite block was sliced into several pieces. The element distribution on the specimen surface was measured by EPMA and an optical microscope, and the gamma activities in the sliced bentonite pieces were measured. All the leachates were sampled and then the gamma and gross alpha activities were measured. By the results, coming from leaching up to about 1.5 years, the fraction of radiocesium released through the compacted bentonite layer from the spent fuel is approximately a hundredth lower than that without the bentonite layer. By the depth profile of the gamma activity on the bentonite, the activity gradually decreased in accordance with more distance from the specimen and the fractional release rates of cesium were nearly the same. On the other hand, the EPMA results indicate that more time and more information is required in order to identify the surface alteration of spent fuel by leaching. This experiment will therefore be continued at least into the year 2006.
Electrochemical Measurements during the Anaerobic Corrosion of Steel
Authors: Smart N.R. , Fennell P.A.H. , Peat R. , Spahiu K. , Werme L. 
Published: May 2011

ABSTRACTIn Sweden, it is proposed that spent fuel should be encapsulated in sealed cylindrical canisters for disposal in a geologic repository. The canisters would consist of a thick ferrous inner container and a copper overpack. If mechanical failure of the copper overpack occurred, allowing water to enter, the ferrous inner container would corrode anaerobically and liberate hydrogen. The rate of hydrogen generation due to the anaerobic corrosion of steel in anoxic groundwater has been measured using barometric cells. The aim of the work presented in this paper was to measure the redox potential, Eh and pH in the presence of anaerobically corroding steel, in a barometric cell. Two specially designed barometric cells were constructed. They were equipped with a silver chloride or calomel reference electrode, a gold Eh electrode, a glass electrode, and a steel electrode. The electrodes were allowed to stabilize in anoxic artificial groundwater and then a mass of pickled steel wire was introduced into the test cell. As the wires were added, the redox potential rapidly became more negative due to the rapid consumption of the residual oxygen. The corrosion potential of the steel was stable and a slow drift in pH was observed. The results are compared with the results of geochemical modelling. Extension of the work to investigate the electrochemical parameters in the presence of dissolved uranium species is discussed.
Evaluation of the Possible Susceptibility of Titanium Grade 7 to Hydrogen Embrittlement in a Geologic Repository Environment
Authors: Greene Charles A. , Henry Alvin J. , Brossia C. Sean , Ahn Tae M. 
Published: May 2011

ABSTRACTTi grade 7 has been selected by the U.S. DOE as the current material of choice for the drip shield in the proposed high level waste (HLW) repository design. Due to the addition of Pd, Ti grade 7 exhibits enhanced resistance to hydrogen embrittlement (HE), yet there is relatively little data on HE of this material. Calculations of hydrogen absorption/recombination, solubility, and free energy of hydride formation in Ti and Pd are presented to qualitatively evaluate Keff, the stress intensity factor for crack propagation induced by hydride formation, of Ti grade 7 in relation to other Ti alloys without Pd. Calculations were performed that show concentration of hydrogen in Ti grade 7 may exceed the critical hydrogen concentration, Hc, where the material becomes embrittled, when accelerated passive dissolution of Ti grade 7 in concentrated Cl- and Cl-+F- solutions as the source of hydrogen is considered.
The Influence of Ligands and Precipitates on the Release of Nuclides from the Near Field under Natural Repository Conditions
Authors: Liu L. , Neretnieks I. 
Published: May 2011

ABSTRACTOnce groundwater intrudes into a damaged canister and wets the spent fuel pellets, radiation emitted from the spent nuclear fuel splits nearby water into oxidizing and reducing species. This may lead to an oxidizing condition near the fuel pellets. As a result, uranium oxide that makes up the fuel matrix will become more soluble, and the incorporated radionuclides will be released more rapidly. The dissolution process is, however, a dynamic one that can be influenced by many factors. Of great importance are the radiation power of the fuel matrix, the concentration of ligands near the fuel surface, and the transport resistance of the near field. Consequently, the escape of nuclides from the damaged canister is dominated mainly by the intrusion of ligands, and the precipitation/dissolution of secondary phases within the fuel rods. To investigate the possible effects of ligands and precipitates, a coupled dissolution and transport model, which includes the barrier effect of the Zircaloy claddings, is developed. The application of the model to a SKB-specified reference scenario indicates that by far the largest fraction of the oxidized uranium will reprecipitate within the canister. This may significantly decrease the fuel surface available for oxidation and the water available for radiolysis. Subsequently, much less fuel matrix will be dissolved and much less of the other nuclides will be released. Simulations further identify that carbonate and silicate have the greatest influences on the formation of secondary phases, and on the release of nuclides, under natural repository conditions.