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[en] We investigate the microcracking mechanisms responsible for Ti3SiC2/SiC joint damage observed at the macroscopic scale after neutron irradiation experiments in detail. A dual-phase microstructural approach to damage and fracture of Ti3SiC2/SiC joints is developed that uses a finely discretized two-phase domain based on a digital image of an actual microstructure involving embedded Ti3SiC2 and SiC phases. The behaviors of SiC and Ti3SiC2 in the domain are described by the continuum damage mechanics (CDM) model reported in Nguyen et al., J. Nucl. Mater., 2017, 495:504–515. This CDM model describes microcracking damage in brittle ceramics caused by thermomechanical loading and irradiation-induced swelling. The dual-phase microstructural model is applied to predict the microcracking mechanisms occurring in a typical Ti3SiC2/SiC joint subjected to heating to 800 °C followed by irradiation-induced swelling at this temperature and cooling to room temperature after the applied swelling has reached the maximum swelling levels observed in the experiments for SiC and Ti3SiC2. The model predicts minor damage of the joint after heating but significant microcracking in the SiC phase and along the boundaries between SiC and Ti3SiC2 as well as along the bonding joint during irradiation-induced swelling and cooling to room temperature. Our predictions qualitatively agree with the limited experimental observations of joint damage at this irradiation temperature.
[en] Defect-free dislocation channel formation has been reported to promote plastic instability during tensile testing via localized plastic flow, leading to a distinct loss of ductility and strain hardening in many low-temperature irradiated materials. In order to study the underlying mechanisms governing dislocation channel width and formation, the channel formation process is modeled via a simple stochastic dislocation-jog process dependent upon grain size, defect cluster density, and defect size. Dislocations traverse a field of defect clusters and jog stochastically upon defect interaction, forming channels of low defect-density. And based upon prior molecular dynamics (MD) simulations and in-situ experimental transmission electron microscopy (TEM) observations, each dislocation encounter with a dislocation loop or stacking fault tetrahedron (SFT) is assumed to cause complete absorption of the defect cluster, prompting the dislocation to jog up or down by a distance equal to half the defect cluster diameter. Channels are predicted to form rapidly and are comparable to reported TEM measurements for many materials. Predicted channel widths are found to be most strongly dependent on mean defect size and correlated well with a power law dependence on defect diameter and density, and distance from the dislocation source. Due to the dependence of modeled channel width on defect diameter and density, maximum channel width is predicted to slowly increase as accumulated dose increases. The relatively weak predicted dependence of channel formation width with distance, in accordance with a diffusion analogy, implies that after only a few microns from the source, most channels observed via TEM analyses may not appear to vary with distance because of limitations in the field-of-view to a few microns. Furthermore, examinations of the effect of the so-called “source-broadening” mechanism of channel formation showed that its effect is simply to add a minimum thickness to the channel without affecting channel dependence on the given parameters.
[en] Molecular dynamics simulations have been performed to investigate oxygen transport in (UxPux-1)0.95Gd0.05O1.975, (UxThx-1)0.95Gd0.05O1.975 and (PuxThx-1)0.95Gd0.05O1.975 between 1000 and 3200 K. Oxygen diffusivity and corresponding activation energies are examined and compared to values for the undoped (UxPux-1)O2, (UxThx-1)O2 and (PuxThx-1)O2 systems where compositions between end members display enhanced diffusivity. Below the superionic transition oxygen diffusivity for the Gd doped systems is orders of magnitude greater compared to their undoped counterparts. But, enhanced diffusivity for doped mixed actinide cation compositions is not observed compared to doped end members. Furthermore, changes in activation energy suggest changes in diffusion regime, which correspond to the creation of thermally activated oxygen defects.
[en] A mono-metallic thermal convection loop (TCL) fabricated from alloy APMT (Fe21Cr5Al3Mo) tubing and filled with 0.025 m long tensile specimens of the same alloy was operated continuously for 1000 h with commercially pure Pb-17 at.%Li (Pb-Li) at a peak temperature of 550 ± 1.5 °C and a temperature gradient of ~116 °C. The resulting Pb-Li flow rate was ~0.0067 m/s. A 1050 °C pre-oxidation treatment (to form an external alumina scale) given to most specimens exposed within the TCL decreased total mass loss by a factor of 3–30 compared to adjacent specimens that were not pre-oxidized. However, all specimens exposed above 500 °C lost mass suggesting that the alumina scale was not entirely stable in flowing Pb-Li at these temperatures. Post-exposure room temperature tensile tests indicated that the mechanical properties of APMT were substantially influenced by extended exposures in the range of 435–490 °C, which caused an increase in yield strength (~65%) and a corresponding decrease in ductility associated with α' embrittlement. Specimens annealed in argon at the same temperature exhibited identical changes without exposure to Pb-Li. In conclusion, scanning transmission electron microscopy revealed Cr-clusters within the microstructure in specimens exposed in the low temperature regions (<490 °C) of the TCL, indicating the formation of α' consistent with the mechanism of α' embrittlement.
[en] Complete text of publication follows: These proceedings contain the papers presented at the 4-day symposium on Microstructural Processes in Irradiated Materials (MPIM) held in San Antonio, Texas, USA on March 3-7, 2013 as part of the 2013 TMS Annual Meeting and Exhibition. The 2013 MPIM symposium was proposed for a wide range of topics involving nuclear structural and fuel materials to reflect the growing nuclear materials area. The symposium was the sixth in a series of symposia that have been held in the TMS Annual Meeting every two years since the first symposium on March 2-6, 2003 in San Diego, California. The majority of the original contributions to this symposium focused on the experimental and modeling studies on defect production processes in irradiation, radiation-induced phase stability and precipitation, mechanical behaviors, and materials processing. The topics on radiation damage processes in nuclear fuels and fuel related materials and irradiation performance of non-metallic materials and fusion materials have become important components of the symposium. The statistics from the contributed abstracts indicated that the high-chromium ferritic-martensitic steels and advanced ODS ferritic alloys are the most studied materials in a variety of topics, followed by the traditional pressurized water reactor materials such as micro-alloyed ferritic reactor pressure vessel steels and austenitic stainless steels. Other well studied materials include nuclear fuel-related materials, such as oxides, uranium alloys, and Zr-alloys, and fusion materials, such as tungsten and silicon carbide. As in the earlier symposia, a strong emphasis was on the connections between the results of modeling studies on point defects, impurity behaviors, damage accumulation and those of advanced experimental characterization for the microstructural evolution in irradiated or processed materials. The MPIM symposium set new records in the numbers of presentations, with a total of 116 presentations, including 83 oral presentations in 8 sessions and 33 posters in the general poster session. Included among the oral presentations were 14 invited talks from representative nuclear materials areas. After a year-long review process 17 papers have been finally accepted and included in this special issue. (authors)
[en] Recent interest in U3Si2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H2O environments is sparse in available literature. The behavior of U3Si2 in H2O containing atmospheres is investigated and presented in a two-part series of articles. This work examines the behavior of U3Si2 following exposure to pressurized H2O at temperatures from 300 to 350 °C. Testing was performed using two autoclave configurations and multiple redox conditions. Use of solid state buffers to attain a controlled water chemistry is also presented as a means to test actinide-bearing systems. Buffers were used to vary the hydrogen concentration between 1 and 30 parts per million H2. Testing included UN, U3Si5, and UO2. Both UN and U3Si5 were found to rapidly pulverize in less than 5 h at 300 °C. Uranium dioxide was included as a control for the autoclave system, and was found to be minimally impacted by exposure to pressurized water at the conditions tested for extended time periods. Testing of U3Si2 at 300 °C found reasonable stability through 30 days in 1–5 ppm H2. However, pulverization was observed following 35 days. The redox condition of testing strongly affected pulverization. Characterization of the resulting microstructures suggests that the mechanism responsible for pulverization under more strongly reducing conditions differs from that previously identified. Hydride formation is hypothesized to drive this transition. In conclusion, testing performed at 350 °C resulted in rapid pulverization of U3Si2 in under 50 h.
[en] The potential for brittle cleavage fracture is a major concern for martensitic stainless steels which are candidates for fusion reactor structural materials. This study attempts to identify for flawed fusion structures the pertinent fracture resistance or failure parameters and the relationships between these parameters and the basic materials properties which govern cleavage fracture. Several procedures for relating test data to failure prediction, including Charpy-V-notch transition temperature referencing and two-parameter interpolation procedures, are considered; and results are discussed with respect to possible research paths for martensitic stainless steel alloy development. (orig.)
[en] Chemical interactions between UO2 fuel and Zircaloy-4 cladding under isothermal and transient temperature conditions up to the melting point of zircaloy (Zry) are described. The tests were conducted in inert gas (1 to 80 bar) with 10 cm long zircaloy cladding specimens filled with UO2 pellets. In the isothermal tests, the annealing temperature varied between 1000 and 17000C and the annealing period between 1 and 150 min. The transient experiments were conducted from 10000C to maximum temperatures of 1400, 1500 and 16000C. The extent of the chemical reaction depends decisively on whether or not good contact between UO2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO2 to form oxygen-stabilized α-Zr(O) and uranium metal. ZrO2 does not form. The uranium reacts with zircaloy low in oxygen to form a (U, Zr) alloy which is liquid above about 11500C and lies between two α-Zr(O) layers. The isothermal UO2/zircaloy reaction obeys a parabolic rate law. The growth of the reaction layers can be represented in an Arrhenius diagram. (orig.)
[en] In the nuclear technology boron is often applied as a neutron absorbing element, especially in the form of the compound boron carbide (B4C). A survey is given on various kinds of products and specific requirements demanded by the practical application of boron containing neutron absorbing materials in and at the core of the reactor, as shielding material for storage and transport of fuel elements. New developments of europium boride (EuB6) as a neutron absorber for fast reactors result in three different types of materials. A review by Murgatroyd and Kelly concerning the state of technology of neutron absorbers has been completed and more recent publications on absorber materials on the base of boron have been cited. (orig.)