bluuebunny/crossref_metadata_2025_split huggingface dataset:
Migration of Cs-137 and Co-60 in the Australian Arid Zone
Authors: Payne T.E. , Harries J.R. , Itakura T. 
Published: May 2011

ABSTRACTBatch adsorption experiments with Cs-137 and Co-60 were undertaken using representative samples of geologic materials from the arid region that has been selected for an Australian low-level waste repository. The results indicate that the pH is the main factor affecting the adsorption of Co-60 but has little influence on the sorption of Cs-137. The ionic strength affects Cs sorption, with a decrease in Kd associated with higher ionic strength. Selective sorption sites on mica and illite control uptake of trace Cs, whereas the high total site availability of smectite is significant when the total Cs is higher (1 mmol/L). The effects of mineralogy on Cs sorption which were observed for these complex materials confirmed previous results reported for pure minerals.
The Solubility of Metallic Selenium under Anoxic Conditions
Authors: Iida Yoshihisa , Yamaguchi Tetsuji , Nakayama Shinichi , Nakajima Tomoko , Sakamoto Yoshiaki 
Published: May 2011

ABSTRACTThe solubility of metallic selenium was measured in a mixture of 0.1M-NaCl and 0.05M-N2H4under anoxic conditions (O2 < 1 ppm) by both oversaturation and undersaturation methods. Equilibrium was attained in 40 days. The aqueous selenium species identified were HSe at pH between 5 and 8, and Se42at pH between 10 and 13, by UV-Vis absorption spectrometry. The solid phase was identified as Se (cr) by X-ray diffraction. The equilibrium constants ofSe(cr) + H+ + 2e- = HSe- logK0 = -6.5±0.5 and4Se(cr) + 2e- = Se42- logK0 = -16.8±0.5were determined. The standard molar free energy of formation of HSe- and Se42-was determined to be (37.1±2.9) and (95.9±2.9) kJ/mol, respectively.
Isolation and Microscopic Characterization of Nuclear Fuel Particles from Two Contaminated Soils of Chernobyl
Authors: Noureddine Ahamdach , Denise Stammose 
Published: May 2011

ABSTRACTNuclear fuel particles were separated, from two different Chernobyl contaminated soils, by sedimentation in bromoform. Their physico-chemical characteristics were studied by scanning electron microscopy. The two soils show the same types of fuel particles: one part of these particles contains U and O and the other part contains U, Zr and O. The structure and the surface morphology of the studied particles are variable and are characteristic of the conditions of their formation. The source term is thus heterogeneous. This heterogeneity has direct consequences on the dissolution of the fuel particles and should be taken into account in mathematical models.
Tritium and Iodide Diffusion through Opalinus Clay
Authors: Llano A. Yllera de , Sainz-Ezquerra M. Mingarro , Gutiérrez M. García 
Published: May 2011

ABSTRACTThe International Mont Terri Project [1, 2] started in 1995 under the patronage of the Swiss National Hydrological and Geological Survey (SNHGS), and has the authorization of the République et Canton du Jura. The underground rock laboratory is located at the northwestern part of Switzerland (Canton Jura), in and beside the reconnaissance gallery of the Mont Terri motorway tunnel, one of the several tunnels of the A16 “Transjurane” motorway. The depth of overburden above the rock laboratory is approximately 300 meters.The project is aimed to investigate the geological, hydrogeological, geochemical and rock mechanical properties of the Opalinus Clay for assessing the feasibility and safety of a repository for radioactive waste placed in this type of host rock. One of the issues under study is radionuclide migration by diffusion through clays. As a part of this investigation, an interlaboratory comparison on small-scale diffusion experiments was carried out by three research laboratories: AEA Technology (UK), SCK·CEN (Belgium) and CIEMAT (Spain). The radionuclides investigated were tritium and iodine. This paper concerns to the methodological approach and results of the experiments undertaken by CIEMAT.The effective diffusion coefficients were measured for tritiated water and iodine (as I-), resulting larger for tritium [(1.7±0.4)×10−11 m2/s] than for iodide [(2.7±0.3)×10−12 m2/s]. The porosity available for diffusion was calculated by using the time-lag method, but some results seemed unrealistic and showed a large variability. In general, tritium exhibited higher values of porosity than iodide (17 to 26% and 12 to 17%, respectively), which were consistent with the anion exclusion affecting the distribution of iodide into the clay pores.
Electrical Conductivity and Structural Properties of Cesium Iron Phosphate Glasses: A Potential Host for Vitrifying Nuclear Waste
Authors: Mogus-Milankovic  , Furic K. , Day D. E. 
Published: May 2011

ABSTRACTThe thermally stimulated current (TSC) and dc conductivity for iron phosphate glasses containing up to 28 mol% Cs2O have been measured in a temperature range from 120 to 400 K. The dc conductivity and activation energy were constant and independent of Cs2O content. With increasing cesium concentration in cesium iron phosphate glasses the slowly moving cesium ions are more tightly bound to the non-bridging oxygen ions and make no measurable contribution to dc conductivity. The dc conduction in these glasses is totally electronic, controlled by electron hopping between iron ions. The ionic conduction is immeasurably small because of the low mobility of the cesium ions. This agreement is reinforced by the excellent chemical durability of the glasses, where the dissolution rate at 90oC changes little with increasing Cs2O content. Raman spectroscopy indicated that the structure of these glasses was composed of predominantly pyrophosphate (P2O7) groups, but the metaphosphate chains (PO3) also existed.
Influence of Platinum-Group Metals on Nuclear Glass Properties: Viscosity, Thermal Stability and Alterability
Authors: Bart F. , Dussossoy J.L. , Fillet C. 
Published: May 2011

ABSTRACTThe fission products resulting from reprocessing of commercial spent fuel are currently vitrified industrially by COGEMA at La Hague. The properties of 21 non-radioactive borosilicate glass samples containing between 4 and 6% of the platinum-group metals (PGM: Pd and Ru) compared with 1.6% in the industrial glass were investigated for chemical composition variations covering the full specification range. After a brief morphological description of the undissolved PGM in the glass, the viscosity variations at temperatures ranging from 1100 to 1200°C are discussed with emphasis on the effects of the particle inclusions on the rheological properties of the glass. Variations in the chemical durability of quenched glass specimens are then discussed. The initial leach rate V0 at 100°C remained near the values obtained for the reference glass. The same tests were repeated on glass heat-treated to obtain maximum crystallization, and the results confirmed that the chemical durability of the glass is practically unaffected by the crystallization observed in this type of glass.
Melted Synthetic Zirconolite-Based Matrices: Effect of Cooling Rate and Heat Treatment on Ceramic Microstructure and Chemical Durability
Authors: Advocat T. , McGlinn P.J. , Fillet C , Leturcq G. , Schuller S. , Bonnetier A. , Hart K. 
Published: May 2011

ABSTRACTThe heat treatment conditions are a key factor in fabricating zirconolite ceramics and glass-ceramics following high-temperature melting. An oxide mixture melted at 1450°C and subsequently heat-treated at 1200°C yielded a glass-ceramic containing crystallized zirconolite–2M. The silica-enriched residual glass represented about 60-70 vol% of the total; the actinide surrogates (Nd, Ce)were equally distributed between the residual glass and the zirconolite crystals. Zirconolite ceramics obtained after melting an oxide mixture at 1600–1700°C consisted of zirconolite, perovskite and rutile. Rapid cooling rates (> 100°#x00B0;··min-1) were obtained by pouring the melt into ingot molds; the resulting zirconolite ceramics were characterized by crystals of zirconolite-2M ranging from 1 to no more than 20 μm. Slow cooling (< 25°C#x00B0;··min-1 produced ceramics with crystals several hundred micrometers long. Despite the microstructural differences, the chemical durability of the zirconolite ceramics was identical. The initial alteration rates r0 were about two orders of magnitude lower than those measured for the residual aluminosilicate glass of the zirconolite glass-ceramics. Moreover, during long-term leach tests at high S/V ratios to obtain advanced degrees of reaction progress, the alteration rates of all the materials diminished by over 3 to 4 orders of magnitude below r0.
Dissolution of Synthetic Brannerite at 90°C
Authors: Zhang Y. , Hart K.P. , Thomas B.S. , Aly Z. , Li H. , Carter M. 
Published: May 2011

ABSTRACTBrannerite, as a minor phase, exists in the pyrochlore-rich titanate ceramic formulations designed for immobilization of surplus weapons Pu. The dissolution of synthetic brannerite was studied at 90°C using static tests in pH 4 solution, deionized water and Finnsjön synthetic groundwater.After 140 days the normalized U release rates into a pH 4 solution and deionized water reach similar values, ~10−2 g m−2 d−1, and are about 2 orders of magnitude higher than those in Finnsjön synthetic groundwater. The normalized Ti release rate into Finnsjön synthetic groundwater is about an order of magnitude higher than those in pH 4 solution and deionized water. The dissolution of brannerite is incongruent in the pH 4 solution and deionized water (preferential release of U over Ti), and nearly congruent in Finnsjön synthetic groundwater. SEM observations of the samples after 140 days in pH 4 solution and deionized water revealed minor surface alteration, in the form of a thin surface layer, probably TiO2, as a result of preferential releases of U in both cases.
Emanation Thermal Analysis Study of Brannerite
Authors: Balek V. , Zelenák V. , Málek Z. , Vance E. R. , Subrt J. 
Published: May 2011

ABSTRACTRadon emanation thermal analysis (ETA) during ”step by step”heating to selected temperatures up to 1200oC in argon and subsequent cooling yielded detailed information on the annealing of radiation damage and macroscopic defects in a natural brannerite. Thermogravimetry showed mass losses in the temperature ranges 230-315, 570-760 and 840-1040oC. Mass spectrometry of the evolvedgases indicated M/Z = 44, i.e. CO2. Crystallization of initially amorphous brannerite was indicated by ETA in the temperature range 900-1020oC, in agreement with the X-ray diffraction results. From the radon diffusion activation energy values determined in ETA, the degree of structural ordering was assessed at different temperatures.
Measured Displacement Energies of Oxygen Ions in Zirconolite and Rutile
Authors: Smith Katherine L. , Cooper Ronald , Colella Michael , Vance Eric R. 
Published: May 2011

ABSTRACTOptical emission spectra in the 300-700 nm range were collected from zirconolite and rutile specimens irradiated with a 3 μs pulsed electron beam using a Febetron 706 variable energy pulsed electronbeam generator. The long-lived emissions (up to microseconds after the electron pulse) consist of broad (halfwidths ~ 100 nm) bands centred around ~400 nm. Over the range 0.2 MeV to 0.6 MeV, the emission intensity per unit dose versus electron beam energy data from the rutile sample showed a single stage dependence on electron beam energy, whereas the zirconolite data suggested a two stage dependence. Rutile has a threshold of 0.23 ½ 0.02 MeV, which gives an Ed value of 39 ½ 4 eV for oxygen. Zirconolite has a threshold of 0.26 ½ 0.02 MeV, which gives an Ed value of 45 ½4 eV for oxygen. These data are discussed in the context of previously measured and calculated Ed values for other oxides.
Influence of Container Base Material (Fe) on SIMFUEL Leaching Behavior
Authors: Quiñones J. , Serrano J.A. , Díaz P.P. , Almazán J.L. Rodríguez , Cobos J. , Esteban J.A. , Martà–nez-Esparza A. 
Published: May 2011

ABSTRACTThe chemical stability of spent fuel will be greatly influenced by the redox potential of the near field. Presence of reductants such as iron is likely to be an important factor to maintain the original integrity of spent fuel. In this work experimental data about the influence of metallic iron (container base material) on SIMFUEL leaching behavior under simulated granite and saline repository conditions is presented. In the presence of iron uranium concentration undergoes a sharp decrease. This is much more noticeable in the experiments performed under initial oxic conditions. The effect of iron on simulated fission products of SIMFUEL is very important for the elements with high redox sensitivity such us molybdenum. On the contrary, strontium remains stable during the entire tests and it seems not be affected by changes in redox potential.
Release of Radionuclides from Spent Fuel Under Repository Conditions: Mathematical Modelling and Preliminary Results
Authors: Cera Esther , Merino Juan , Bruno Jordi 
Published: May 2011

ABSTRACTIn the framework of the Enresa 2000 PA exercise and as a continuation of the developments made during SR 97, we have developed a conceptual and numerical model to calculate the release of radionuclides from spent fuel under repository conditions. The model includes both thermodynamic and kinetic considerations. Hence, although certain radionuclides are solubility controlled, for other radionuclides their release is governed by kinetic processes such as radiolytically promoted oxidative dissolution of the matrix and the associated water turnover inthe gap. The fluxes of selected radionuclides are calculated as an indication of the relative importance of the various processes considered to define source term concentrations in the performance assessment of the spent fuel repository.
A New Approach to the RN Source Term for Spent Nuclear Fuel under Geological Disposal Conditions
Authors: Poinssot Christophe , Jegou Christophe , Toulhoat Pierre , Piron Jean-Paul , Gras Jean-Marie 
Published: May 2011

ABSTRACTUnder the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before coming in contact with water. Under these circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides. The major mechanisms are the production of a significant volume of He within the UO2 lattice, the accumulation of irradiation defects due to the low temperature which avoids any annealing, the slow migration of radionuclides (RN) within the system by (i) the α self-irradiation-induced athermal diffusion and (ii) locally the building-up of internal mechanical stresses which could turn the pellets into powder. However, the current RN source terms for SF have never accounted for this evolution as they are based on existing knowledge of the fresh SF. Two major mechanisms were considered, the leaching of the readily available fraction (one which was supposed to be instantly accessible to water), and the release of RN through alteration of the UO2 grains. We are now proposing a new RN source term model based on a microscopic description of the system in order to also take account of the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.
Mechanical Properties of Oxides Formed by Anaerobic Corrosion of Steel
Authors: Smart N.R. , Bond A.E. , Crossley J.A.A. , Lovegrove P.C. , Werme L. 
Published: May 2011

ABSTRACTIn Sweden, it is proposed that spent nuclear 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, there would be a build up of ferrous corrosion product, which could induce stresses in the outer copper canister. This paper describes an apparatus, the ‘stress cell’, which was designed to measure the expansion caused by the anaerobic corrosion of steel under compressive loads. The apparatus consisted of a stack of steel and copper discs, which were immersed in simulated anoxic groundwater. A system of levers amplified the change in height of the stack, and the displacement was measured using sensitive transducers. Three cells were set up; two contained alternate mild steel and copper discs, and the third, a control cell, consisted of alternate stainless steel and copper discs. A slight contraction of the control cell was observed but no expansion was measured in the mild steel - copper cells.In parallel, coupons of mild steel and cast iron were corroded in anoxic, artificial groundwater at 50°C and 80°C for several months. The coupons were examined by atomic force microscopy (AFM) to determine the mechanical properties and the structure of the corrosion product films, and X-ray photoelectron spectroscopy (XPS) to identify the chemical composition of the film.
Influence of Oxygen Concentration and Alkalinity on the Hydrogen Gas Generation by Corrosion of Carbon Steel
Authors: Fujiwara A. , Yasutomi I. , Fukudome K. , Tateishi T. , Fujiwara K. 
Published: May 2011

ABSTRACTA repository for low and intermediate level radioactive wastes will contain a great deal of carbon steel as reinforcement materials and waste containers. These steel components are expected to generate hydrogen gas due to reductive corrosion of carbon steel under repository condition. In this study, we have investigated the influence of the environmental factors such as oxygen concentration and pH in the solution on the rate of the gas generation. This rate was measured with a gas flow type test system under the controlled oxygen concentration, which simulates the redox conditions from initial to final environment in a repository. In addition, the corrosion mechanism was examined through the electrochemical tests.As a result, it has been found that pH in the solution and the oxygen concentration considerably affected the generation rate of hydrogen gas in various way. As to the influence of pH of the solution, it has been found that the gas generation rate increased with the increase of pH between 11.8 and 14. Further, as to the influence of the oxygen concentration, when the oxygen concentration in the blowing nitrogen gas was 2 ppm or more, the gas generation rate increased with the oxygen concentration. This trend was possibly caused by local decrease of pH due to localized corrosion. The gas generation rate was very low when the oxygen concentration was in the passive region. The rate was equal to 0.01 μm/y or less if the corrosion occurs with an equation: 3Fe+4H2O → Fe3O4 + 4H2 ↑ to 0.01 *m imcro;/y (hereafter we call this the equivalent corrosion rate). However the gas generation rate increased when the oxygen concentration was controlled to the extent that sufficient passive film cannot be produced and thus it shifts from passive state to active state.
Study of the Consequences of Secondary Water Radiolysis within and Surrounding a Defective Canister
Authors: Liu Jinsong , Strömberg Bo , Neretnieks Ivars 
Published: May 2011

ABSTRACTA model has been developed to study the effects of secondary water radiolysis caused by dispersed radionuclides in a bentonite buffer surrounding a copper canister. The secondary radiolysis is the radiolysis caused by radionuclides that have been released from the spent fuel and are present either as solutes in the pore-water, as sorbed species on the surface of other minerals, or as secondary minerals. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The dissolution of the spent fuel is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases have been considered with the purpose to illustrate the behaviors of both conservative and non-conservative nuclides. The nuclides that are most relevant are those expected to be the dominant α-emitters in the long-term (e.g. 239Pu, 240Pu, 241Am). In the first case it is assumed that there is no precipitation of secondary minerals of the relevant radionuclides inside the canister. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The radionuclide released to the surrounding buffer is then predicted using a mass balance model. The modelling results show that in both cases, the spent fuel will not be oxidized at a rate significantly faster compared to the case where secondary radiolysis is completely neglected. In the first case, however, a large domain of the near-field can be oxidized due to a much faster depletion of reducing minerals in the buffer, compared to the case where secondary radiolysis is neglected. In the second case, the effects of the secondary water radiolysis will be quite limited.
Coupled Thermo-Hydro-Geochemical Models of Engineered Barrier Systems: The Febex Project
Authors: Samper J. , Juncosa R. , Navarro V. , Delgado J. , Montenegro L. , Vázquez A. 
Published: May 2011

ABSTRACTFEBEX (Full-scale Engineered Barrier EXperiment) is a demonstration and research project dealing with the bentonite engineered barrier designed for sealing and containment of waste in a high level radioactive waste repository (HLWR). It includes two main experiments: an situ full-scale test performed at Grimsel (GTS) and a mock-up test operating since February 1997 at CIEMAT facilities in Madrid (Spain) [1,2,3]. One of the objectives of FEBEX is the development and testing of conceptual and numerical models for the thermal, hydrodynamic, and geochemical (THG) processes expected to take place in engineered clay barriers. A significant improvement in coupled THG modeling of the clay barrier has been achieved both in terms of a better understanding of THG processes and more sophisticated THG computer codes. The ability of these models to reproduce the observed THG patterns in a wide range of THG conditions enhances the confidence in their prediction capabilities. Numerical THG models of heating and hydration experiments performed on small-scale lab cells provide excellent results for temperatures, water inflow and final water content in the cells [3]. Calculated concentrations at the end of the experiments reproduce most of the patterns of measured data. In general, the fit of concentrations of dissolved species is better than that of exchanged cations. These models were later used to simulate the evolution of the large-scale experiments (in situ and mock-up). Some thermo-hydrodynamic hypotheses and bentonite parameters were slightly revised during TH calibration of the mock-up test. The results of the reference model reproduce simultaneously the observed water inflows and bentonite temperatures and relative humidities. Although the model is highly sensitive to one-at-a-time variations in model parameters, the possibility of parameter combinations leading to similar fits cannot be precluded. The TH model of the “in situ” test is based on the same bentonite TH parameters and assumptions as for the “mock-up” test. Granite parameters were slightly modified during the calibration process in order to reproduce the observed thermal and hydrodynamic evolution. The reference model captures properly relative humidities and temperatures in the bentonite [3]. It also reproduces the observed spatial distribution of water pressures and temperatures in the granite. Once calibrated the TH aspects of the model, predictions of the THG evolution of both tests were performed. Data from the dismantling of the in situ test, which is planned for the summer of 2001, will provide a unique opportunity to test and validate current THG models of the EBS.