bluuebunny/crossref_metadata_2025_split huggingface dataset:
Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel
Authors: Fortner Jeffrey A. , Kropf A. Jeremy , Jerden James L. , Cunnane James C. 
Published: October 2011

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.
The BRAG and GM2003 Models for Glass Dissolution
Authors: Aertsens Marc 
Published: October 2011

AbstractThe GM2003 model extends the r(t) glass dissolution model with water diffusion through the diffusion layer. Boron and alkali diffusion through the diffusion layer is described by introducing a retention factor Kd,i between boron/alkali and water in the diffusion layer. Introducing a boron/alkali diffusion coefficient, the BRAG model describes boron/alkali diffusion in the diffusion layer as well. It is shown that both models are consistent with each other and an expression is derived for the boron/alkali diffusion coefficient (BRAG) as a function of both parameters of GM2003: the retention factor Kd,i and the water diffusion coefficient DH2O in the diffusion layer. From dissolution data only, it is possible to fit the value for the boron/alkali diffusion coefficient in the diffusion layer but due to correlations the individual values of both parameters Kd,i and DH2O of GM2003 cannot be determined. From theoretical considerations follows that the Kd,i value for boron/alkali should be slightly larger than 0.1 kg/liter. A user friendly code for the BRAG model allows automatic fits of glass dissolution data in water.
Microstructural Characterization of U Coprecipitated Phases Formed in Bentonic-Granitic Groundwater and under Anoxic Conditions
Authors: Quinones Javier , Iglesias Eduardo , Cobo Jose M. , Martinez Esparza Aurora , Gomez de Salazar Jose Maria 
Published: October 2011

AbstractFor improving the accuracy of the performance assessment studies related to the spent fuel safety under storage conditions it is necessary to develop a new matrix alteration model. These models must be based on laboratory experiences and they should be capable to extrapolate to storing environmental conditions. Most of recently models developed included the oxidation and dissolution process of the spent fuel matrix, but the influence of a possible process of secondary phase formation over the spent fuel surface is not taken into account yet. This is a key process that could produce a reduction of the matrix dissolution rate, radiation shielding behaviour; however, the surface precipitation of the secondary phase could induce a localized corrosion process, which in this case dissolution rate of the spent fuel would be increased. This paper is focussed on microstructural characterization of secondary phases formed in coprecipitation experiments performed under anoxic conditions in granitic-bentonitic simulated groundwater. In order to simulate the influence of the container material, the coprecipitation experiments were performed in absence and presence of iron powder. The solid phases formed were characterized using the following techniques: XRD; SEM-EDX and TEM-EDX. The XRD diffraction pattern showed that under anoxic conditions a mixture of phases were obtained (sodium and potassium uranate and schoepite), whereas uranate phases were detected when only iron was present. The characterization study indicates that the U secondary phase formed (under reducing conditions and presence of iron powder) growth from iron surface. The crystal size of the secondary phase is independent of the presence of iron powder (and it is always less than 3 μm). Furthermore, the microstructural study showed the growing of U phases over iron powder.
Development of a Phosphate Ceramic as a Host for Halide-contaminated Plutonium Pyrochemical Reprocessing Wastes
Authors: Metcalfe Brian , Fong Shirley , Gerrard Lee , Donald Ian , Strachan Denis , Scheele Randall 
Published: October 2011

AbstractThe presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable, leach resistant form. AWE has developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process using actinide-doped Type I waste (essentially a chloride-based waste) were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined using a modified MCC-1 test at 40°C/28 days were 12 x 10-6 g.m-2 and 2.7 x 10-3 g.m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (an oxyhydroxide-based waste) is currently underway at PNNL.Differences in the ionic state of plutonium in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Early investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste showed some apparent differences in the phase assemblages of the surrogate and actinide-based products. However XRD examination of the products at higher resolution has demonstrated there is no significant difference and that for this work HfO2 is a suitable surrogate for PUO2.
Modeling of Spent Fuel Oxidation at Low Temperature
Authors: Poulesquen Arnaud , Desgranges Lionel , FERRY Cécile 
Published: October 2011

AbstractDuring dry storage, the oxidation of the spent fuel in case of cladding and container failure (accidental scenario) could be detrimental for further handling of the spent fuel rod and for the safety of the facilities. Depending on whether the uranium dioxide is under the form of powder or pellet, irradiated or unirradiated, the weight gain curves do not present the same shape. To account for these different behaviours, two models have been developed. Firstly, the oxidation of unirradiated powders has been modelled based on the coexistence, during the oxidation, of two intermediate products, U4O9 and U3O7. The comparison between the calculation and the literature data is good in terms of weight gain curves and chemical diffusion coefficient of oxygen within the two phases. Secondly, the oxidation of spent fuel fragments is approached by a convolution procedure between a grain oxidation model and an empirical parameter which represents the linear oxidation speed of grain boundary or an average distance able to cover the entire spent fuel fragment. This procedure of calculation allows in one hand to account for the incubation period noticed on unirradiated pellets or spent fuel and in another hand to link the empirical parameter to physical as porosity, cracks or linear power, or operational parameters such as fission gas release (FGR) respectively. A comparison of this new modelling with experimental data will be proposed.
EELS Spectrum Imaging and Tomography Studies of Simulated Nuclear Waste Glasses
Authors: Yang Guang , Saghi Zineb , Xu Xiaojing , Hand Russell , Möbus Günter 
Published: October 2011

AbstractElectron energy loss spectroscopy (EELS) has been widely used in analysis of ceramic, minerals and semiconductors. It can provide the unique advantage of sensitivity to composition, coordination and valency, while providing nano-scale spatial resolution. Due to the electron irradiation sensitivity, EELS studies on glass are quite rare. However, using special care in glass-composition selection and by recording time series of EELS-spectra to monitor any structural changes, EELS then becomes a very useful technique to study glass-chemistry and glass structure with the highest spatial resolution of all chemically sensitive techniques. Alkali borosilicate glasses (ABS) doped with Cr2O3 (2 mol%), CeO2 (4 mol%) and ZrO4 (4 mol%) were melted, cooled and annealed in the context of simulating radionuclide immobilisation glasses. Precipitates with diameter in the range of ∼20 nm to ∼200 nm were found homogeneously distributed in the glass. In preliminary studies we measured the oxidation state in the glass and in precipitates via evaluation of the Ce-M-edge double white line ratio, confirming the crystals as Ce(IV) oxide. We also found Boron K-edge ELNES spectra as a sensitive signal to boron coordination (N4 = BO4/(BO3+BO4)), which is found at N4 ratios of around 35-45%.In the present study we have extended this work to the acquisition of dense line scans (spectrum imaging) across an area of interest on our composite glass with emphasis on three questions:(i)	Change of cerium valence upon transition from the glass to the precipitates?(ii)	Possible oscillation of boron N4-ratio in the glass within a line scan as a consequence of random medium-order related glass fluctations when scanning the <5nm probe across a seemingly homogeneous area?(iii)	N4-changes in the immediate glass layer surrounding the precipitates?Our ongoing research has produced results which prove a systematic change of the Ce oxidation state from precipitates (+IV) to glass (mixed +III/+IV). It is found that boron coordination is constant within the sensitivity given by signal-to-noise in the glass matrix. A change of the Boron K-edge fine structure upon crossing a nanoparticle during one particular linescan has been observed, which could hint to a coordination change of the glass layer surrounding the particle. Other linescan measurements have not shown any change and acquisition of an enlarged better statistic is work in progress.The ABS-glass with distributed nanoparticles provides an ideal example of a 3D nanocomposite. The only proven technique for the three-dimensional reconstruction of such materials on the nanoscale is electron tomography. We present the first results of a 3D reconstructed nuclear waste glass by using a tilt series of ADF STEM images covering a glass fragment of 2000nm field of view containing several tens of nanoparticles distributed around its volume.
Tc and Re Behavior in Borosilicate Waste Glass Vapor Hydration Tests
Authors: McKeown David A. , Buechele Andrew C. , Lukens Wayne W. , Shuh David K. , Pegg Ian L. 
Published: October 2011

AbstractTechnetium (Tc), found in some nuclear wastes, is of particular concern with regard to long-term storage, because of its long-lived radioactivity and high mobility in the environment. Tc and rhenium (Re), commonly used as a non-radioactive surrogate for Tc, were studied to assess their behavior in borosilicate glass under hydrothermal conditions in the Vapor Hydration Test (VHT). X-ray absorption spectroscopy (XAS) and scanning electron microscopy (SEM) measurements were made on the original Tc- and Re-containing glasses and their corresponding VHT samples, and show different behavior for Tc and Re under VHT conditions. XAS indicates that, despite starting with different Tc(IV) and Tc(VII) distributions in each glass, the VHT samples have 100% Tc(IV)O6 environments. SEM shows complete alteration of the original glass, Tc enrichment near the sample surface, and Tc depletion in the center. Perrhenate (Re(VII)O4−) is dominant in both Re-containing samples before and after the VHT, where Re is depleted near the VHT sample surface and more concentrated toward the center.
Nuclear Waste Disposal in Deep Geological Formations: What are the Major Remaining Scientific Issues?
Authors: Toulhoat Pierre 
Published: October 2011

AbstractAfter 30 years of research in the field of nuclear waste disposal, the remaining scientific issues are described. They include long-term monitoring during the reversible period, early coupled processes during the thermal phase, effect of hydrogen generated by corrosion, and volume expansion. The long term behavior of waste packages has been thoroughly investigated, but controversial issues have still to be resolved. In situ investigations concerning the speciation and migration of radionuclides are needed. Concerning the developement of future nuclear systems, a roadmap on waste management issues has to be established.
A High-Performance Corrosion-Resistant Iron-Based Amorphous Metal - The Effects of Composition, Structure and Environment on Corrosion Resistance
Authors: Farmer Joseph Collin , Haslam J. , Day D. , Lian T. , Saw C. , Hailey P. , Choi J-S. , Rebak R. , Yang N. , Bayles R. , Aprigliano L. , Payer J. , Perepezko J. , Hildal K. , Lavernia E. , Ajdelsztajn L. , Branagan D. , Beardsley B. 
Published: October 2011

ABSTRACTThe passive film stability of several Fe-based amorphous metal formulations have been found to be comparable to that of high-performance Ni-based alloys, and superior to that of stainless steels, based on electrochemical measurements of the passive film breakdown potential and general corrosion rates. Chromium (Cr), molybdenum (Mo) and tungsten (W) provide corrosion resistance; boron (B) enables glass formation; and rare earths such as yttrium (Y) lower critical cooling rate (CCR). The high boron content of this particular amorphous metal also makes it an effective neutron absorber, and suitable for criticality control applications, as discussed in companion publications. Corrosion data for SAM2X5 (Fe49.7Cr17.7Mn1.9Mo7.4W1.6B15.2C3.8Si2.4) is discussed here.
A Review of 25 years of Corrosion Studies on HLW Container Materials at the CEA
Authors: Helie Max 
Published: October 2011

ABSTRACTThe Commissariat à l'Énergie Atomique (CEA, French Atomic Energy Commission) has been involved in researches on nuclear waste management for more than 25 years. One of the key issues is the prediction of the long term behavior and aging of the High Level Waste (HLW) containers in order to develop concepts that will ensure the confinement of the activity over extremely long periods of time.Preliminary studies were carried out on two concepts, one of a thin "corrosion resistant" container made of titanium or nickel base alloy, and the other on a thick "corrosion allowance" container made or carbon steel.The results of these experiments showed that the "corrosion resistant" concept led to a high uncertainty on the development and propagation rate of localized forms of corrosion, and the concept of geological disposal in an argillaceous host formation of thick waste containers made of carbon steel was chosen as the reference. Studies are still in progress to better assess the corrosion mechanisms relevant to this situation in order to provide reliable models for the long term prediction of the containers corrosion behavior.
Cermet Spent Nuclear Fuel Casks and Waste Packages
Authors: Forsberg Charles W. , Dole Leslie R. 
Published: October 2011

ABSTRACTMultipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can outperform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal-conductivity cask.A multiyear, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports.
Dissolution Testing of a Metallic Waste Form in Chloride Brine
Authors: Janney Dawn E. 
Published: October 2011

AbstractArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel from the Experimental Breeder Reactor II (EBR-II). One waste stream from this process consists of a metal waste form (MWF) whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr) and whose microstructure is a eutectic intergrowth of iron solid solutions and Fe-Zr-Cr-Ni intermetallics. This paper reports scanning electron microscope (SEM) observations of corrosion products formed during static immersion tests in which coupons of surrogate MWF containing 10 wt% U (SS-15Zr-10U) were immersed in solutions with nominal pH values of 3 and 4 and 1000 ppm added chloride for 70 days at 50 °C. Although the majority of the surface areas of the coupons appear unchanged, linear areas with localized corrosion products apparently consisting of porous materials overlying corrosion-product-filled channels formed on both coupons, cross-cutting phase boundaries in the original eutectic microstructures. Many of the linear areas intersected the sample edge at notches present before the tests or followed linear flaws visible in pre-test images. Compositions of corrosion products differed significantly from the bulk composition, and the maximum observed concentration of U exceeded that reported in actinide-bearing phases in uncorroded surrogate MWF samples with comparable concentrations of U.
Cesium Sorption Rate on Non-crushed Rock Measured by a New Apparatus based on a Micro-channel-reactor Concept
Authors: Okuyama Keita , Sasahira Akira , Noshita Kenji , Ohe Toshiaki 
Published: October 2011

AbstractSince nuclide migration through rock mediums is an extremely slow process, experimental effort to evaluate the barrier performance of geologic disposal such as the diffusion coefficient (De) and the distribution coefficient (Kd) requires relatively long testing periods and chemically stable conditions. We have developed a fast method to determine both De and Kd by using a non-crushed rock sample. In this method, a fluoroplastic plate with a micro channel (10-200-μm depth) is placed just beneath a rock-sample plate, and a radionuclide solution is injected into the channel at constant rate. A part of radionuclide diffuses into the rock matrix and/or adsorb on the rock surface. The difference between the inlet and outlet radionuclide flux is simply related to the apparent diffusion coefficient (Da) of the rock sample. In this study, we estimated Kd of Cs for granite by using the equilibrium model, finding that Kd decreased with increasing flow rate. This dependence of Kd on flow rate implies the state of sorption equilibrium. The adsorption and desorption curves of 134Cs were thus measured, and the rate constants for both processes were obtained by adopting a first-order rate law. The rate constants of sorption (k+) and desorption (k-) were obtained as a function of flow velocity; constant values of both were observed. Kd was calculated from k+/ k- and then compared with that determined by conventional batch sorption method using a crushed rock sample. The Kd values determined by the present and conventional methods are in good accordance; however, the testing periods for each method are very different, namely, 1 day and 7 days, respectively.
Geochemical Processes at the Carbon steel/bentonite Interface in Repository Conditions
Authors: Torres Elena , Turrero María Jesús , Martin Pedro Luis 
Published: October 2011

AbstractThe Deep Geological Repository (DGR) is currently the most accepted management option for the isolation of high level radioactive wastes. The DGR is based on a multibarrier system, which will limit releases of mobile radionuclides to the biosphere. In the design of the repository the spent fuel is encapsulated in canisters of carbon-steel. The space between the canister and the host geological formation will be filled with bentonite buffer clay. Under the prevailing conditions in a DGR, both localized and generalized corrosion phenomena are possible.Corrosion of the canister will result in formation of solid and gaseous corrosion products, which can influence the behaviour of both the canister and the bentonite. Many studies have been carried out in order to improve the knowledge on the reactivity of these barriers. Most of them have focused on the mineralogical alteration of the bentonite as a function of temperature, time, iron/clay and liquid/rock ratio in batch conditions. The aim of this study is to provide experimental evidences, at repository conditions, on chemical and mineralogical changes during the canister-compacted bentonite interaction: determination of secondary minerals and their alteration reactions, the advance of the corrosion front in the compacted bentonite, and changes in porosity, permeability and cation exchange capacity.