Zircon (ZrSiO{sub 4}) is an actinide host phase in vitreous ceramic nuclear waste forms and a potential host phase for the disposition of excess weapons plutonium. In the present work, the effects of 800 and 900 keV electron and 1 MeV Ne{sup +} irradiations on the structure of single crystals of ZrSiO{sub 4} have been investigated. The microstructural evolution during the irradiations was studied in situ using a high-voltage electron microscope interfaced to an ion accelerator at Argonne National Laboratory. The results indicate that electron irradiation at 15 K cannot amorphize ZrSiO{sub 4} even at fluences an order of magnitude higher than that required for amorphization by 1.5 MeV Kr{sup +} ions. However, the material is readily amorphized by 1 MeV Ne{sup +} irradiation at 15 K. The temperature dependence of this amorphization is discussed in light of previous studies of radiation damage in ZrSiO{sub 4}.
Multiscale approaches are developed to build more physically based kinetic and mechanical mesoscale models to enhance the predictive capability of fuel performance codes and increase the efficiency of the development of the safer and more innovative nuclear materials needed in the future. Atomic scale methods, and in particular electronic structure and empirical potential methods, form the basis of this multiscale approach. It is therefore essential to know the accuracy of the results computed at this scale if we want to feed them into higher scale models. We focus here on the assessment of the description of interatomic interactions in uranium dioxide using on the one hand electronic structure methods, in particular in the density functional theory (DFT) framework and on the other hand empirical potential methods. These two types of methods are complementary, the former enabling to get results from a minimal amount of input data and further insight into the electronic and magnetic properties, while the latter are irreplaceable for studies where a large number of atoms needs to be considered. We consider basic properties as well as specific ones, which are important for the description of nuclear fuel under irradiation. These are especially energies, which are the mainmore » data passed to higher scale models. We limit ourselves to uranium dioxide.« less
The titanium alloy Ti6Al4V is widely used in accelerator facilities as beam windows, which are exposed to high cycle compressive stress waves from intense pulsed proton beams. Such materials interacting with the beam are subject to various forms of radiation damage, which can adversely affect their endurance limit. However, no fatigue data is currently available for high energy proton irradiated titanium alloy. Due to limitations in proton irradiation facilities, only miniature flat samples can be used for irradiation. To address this issue, we have developed a custom-made bend fatigue tester at Fermilab specifically for testing proton irradiated titanium alloy. In this presentation, we report on the fatigue test results obtained from this custom-fatigue tester using a non-ASTM standard specimen design. We plan to validate these sparse fatigue data with ASTM standard samples using standard fatigue samples. Furthermore, we have modified another commercial bend fatigue tester to accommodate miniature samples, and discuss some inherent deficiencies of the commercial fatigue tester to test miniature samples. To overcome this issue, a new fixture design is presented, which enables satisfactory fatigue testing on miniature samples over long periods. Finally, we present an upgrade to the custom-fatigue tester, featuring this new fixture design.
Zirconia is viewed as a material of exceptional resistance to amorphization by radiation damage, and consequently proposed as a candidate to immobilize nuclear waste and serve as an inert nuclear fuel matrix. Here, we perform molecular dynamics simulations of radiation damage in zirconia in the range of 0.1-0.5 MeV energies with full account of electronic energy losses. We find that the lack of amorphizability co-exists with a large number of point defects and their clusters. These, importantly, are largely isolated from each other and therefore represent a dilute damage that does not result in the loss of long-range structural coherence and amorphization. We document the nature of these defects in detail, including their sizes, distribution and morphology, and discuss practical implications of using zirconia in intense radiation environments.
We have studied the effect of the yttria content on the paramagnetic centers in electron-irradiated yttria-stabilized zirconia (ZrO2: Y3+) or YSZ. Single crystals with 9.5 mol % or 18 mol % Y2O3 were irradiated with electrons of 1.0, 1.5, 2.0, and 2.5 MeV. The paramagnetic center production was studied by X-band electron paramagnetic resonance (EPR) spectroscopy. The same paramagnetic centers were identified for both chemical compositions, namely two electron centers, i.e., (i) F+-type centers (involving singly ionized oxygen vacancies), and (ii) so-called T centers (Zr3+ in a trigonal symmetry site), as well as hole-centers. A strong effect is observed on the production of hole-centers that is strongly enhanced when doubling the yttria content. However, no striking effect is found on the electron centers (except the enhancement of an extra line associated with the F+-type centers). It is concluded that hole-centers are produced by inelastic interactions, whereas F+-type centers are produced by elastic collisions with no effect of the yttria content on the defect production rate. In the latter case, the threshold displacement energy (Ed) of oxygen is estimated from the electron-energy dependence of the F+-type center production rate, with no significant effect of the yttria content on Ed. An Ed value larger than 120 eV is found. This is supported by classical molecular dynamics (MD) simulations with a Buckingham-type potential that show Ed values for Y and O are likely to be in excess of 200 eV. Due to the difficulty in displacing O or Y atoms, the radiation-induced defects may alternatively be a result of Zr atom displacements for Ed = 80 ± 1 eV with subsequent defect rearrangement.
Cover Photograph: Geometrical tolerance of extruded glass tubes is an important parameter for applications such as mass spectrometry.Glass flow effects during extrusion such as die swell and tapering can affect the tolerances along the length of extruded tubes.These effects are evaluated via measurements of the internal and external diameter of the extruded tubes using both optical profilometry and microscopy.The figure shows cross-sectional images of precision manufactured glass tubes, from which diameter measurements are taken.The top figures are optical microscope images and the bottom figures are optical profiler images, whereas the left figures demonstrate measurements on cleaved samples and the figures on the right demonstrate samples which have been cut with a diamond saw.High precision extrusion of glass, tubes, Kalnins et al.
We have performed molecular dynamics simulation of displacement events on silicon and carbon sublattices in silicon carbide for displacement doses ranging from 0.005 to 0.5 displacements per atom. Our results indicate that the displacement threshold energy is about 21 eV for C and 35 eV for Si, and amorphization can occur by accumulation of displacement damage regardless of whether Si or C is displaced. In addition, we have simulated defect production in high-energy cascades as a function of the primary knock-on atom energy and observed features that are different from the case of damage accumulation in Si. These systematic studies shed light on the phenomenon of non-ionizing energy loss that is relevant to understanding space radiation effects in semiconductor devices.
