An investigation of the internal friction of manganese-copper alloys

Internal friction is defined as the ability of a vibrating solid to convert its mechanical energy of vibration into heat, even when completely isolated from its surroundings. (1) The most familiar manifestation of internal friction is the damping of a freely vibrating body, such as a torsional pendulum. Another example would be the increased width of the resonance peak observed when a nonelastic body is forced to vibrate over a spectrum of frequencies.(2) Internal friction manifests itself in numerous ways, and we will here be concerned with the first example cited and the potency of data gained in studying the decay of torsional vibration. During the past ten years internal friction has become an increasingly prominent research topic among physicists, metallurgists, and engineers. To the engineer internal friction manifests itself as high damping in alloys, which he can machine into mechanical components having the ability to abate unwanted and destructive modes of vibration. In turn, the metallurgist endeavors to selectively heat treat, age, or alloy various metallic elements and thus produce alloys bearing efficient damping mechanisms. In conjunction with this vast research program, the physicist looks to internal friction for information about the basic structure of solid matter and the laws that govern its dynamic behavior. Interest in internal friction has been shown at the Bureau of Mines in Rolla with regard to the damping capacity of manganese-copper alloys. Extensive investigations by the Bureau of Mines have brought to light the valuable qualities of this alloy as an engineering material, and at the same time have presented data revealing the structural nature of the alloy in various states. A need was felt for information that could be obtained by a study of the behavior of manganese-copper alloys in torsional vibration at very low stress levels. Emphasis was placed on the variation of internal friction and dynamic rigidity with temperature. Thus an investigation was proposed which entailed the following: Design and construction of a Ke type pendulum for measurements under vacuum or inert atmospheres and elevated temperatures. Adaptation of that instrument to the measurement of internal friction and dynamic rigidity of manganese-copper alloys under conditions of varying temperature and reduced pressure. Design and construction of equipment for heat treating manganese-copper wires in the solid solution range, and the development of a technique for rapidly quenching the wire specimens without bending them. Use of this equipment to prepare straight and unoxidized wire specimens (1.3 inches long by 1/32-inch in diameter) of two compositions (85 per cent manganese-15 per cent copper and 75 per cent manganese-25 per cent copper), as quenched from the α -solid solution region, in an attempt to retain the solid solution structure at room temperature. Measurement of the internal friction and dynamic rigidity of these specimens as a function of temperature, at low stress levels and low frequencies. Development of aging techniques for annealing these specimens to produce α -manganese precipitation. Investigation of the effect of the precipitated α -manganese on the internal friction and dynamic rigidity of these alloys. Correlation of the data thus obtained to establish a mechanism for the vibration damping encountered in these alloys --Abstract, pages 1-3

Intrinsic point defects and volume swelling in ZrSiO4 under irradiation

The effects of high concentration of point defects in crystalline ZrSiO4 as originated by exposure to radiation, have been simulated using first principles density functional calculations. Structural relaxation and vibrational studies were performed for a catalogue of intrinsic point defects, with different charge states and concentrations. The experimental evidence of a large anisotropic volume swelling in natural and artificially irradiated samples is used to select the subset of defects that give similar lattice swelling for the concentrations studied, namely interstitials of O and Si, and the anti-site Zr(Si), Calculated vibrational spectra for the interstitials show additional evidence for the presence of high concentrations of some of these defects in irradiated zircon.Comment: 9 pages, 7 (color) figure

Storing Waste in Ceramic

Not all the nuclear waste destined for Yucca Mountain is in the form of spent fuel. Some of it will be radioactive waste generated from the production of nuclear weapons. This so-called defense waste exists mainly as corrosive liquids and sludge in underground tanks. An essential task of the U.S. high-level radioactive waste program is to process these defense wastes into a solid material--called a waste form. An ideal waste form would be extremely durable and unreactive with other repository materials. It would be simple to fabricate remotely so that it could be safely transported to a repository for permanent storage. What's more, the material should be able to tolerate exposure to intense radiation without degradation. And to minimize waste volume, the material must be able to contain high concentrations of radionuclides. The material most likely to be used for immobilization of radioactive waste is glass. Glasses are produced by rapid cooling of high-temperature liquids such that the liquid-like non-periodic structure is preserved at lower temperatures. This rapid cooling does not allow enough time for thermodynamically stable crystalline phases (mineral species) to form. In spite of their thermodynamic instability, glasses can persist for millions of years. An alternate to glass is a ceramic waste form--an assemblage of mineral-like crystalline solids that incorporate radionuclides into their structures. The crystalline phases are thermodynamically stable at the temperature of their synthesis; ceramics therefore tend to be more durable than glasses. Ceramic waste forms are fabricated at temperatures below their melting points and so avoid the danger of handling molten radioactive liquid--a danger that exists with incorporation of waste in glasses. The waste form provides a repository's first line of defense against release of radionuclides. It, along with the canister, is the barrier in the repository over which we have the most control. When a waste form is designed, the atomic environment of the radionuclides is chosen to maximize chemical durability. Elements such as zirconium and phosphorus can be included in the waste form that react with and make some radionuclides less soluble and therefore less likely to be released. The long-term performance assessment of radionuclide containment requires the development of models for each part of the barrier system. It is almost certainly easier to model the corrosion and alteration of waste forms than it is to develop coupled hydrologic, chemical, and geophysical models of radionuclide transport away from a repository. Therefore, much time and effort has been spent optimizing the chemical durability of both glass and ceramic waste forms for radionuclide containment. This has not been an easy task. Three problems in particular posed the greatest challenges. The first is that radionuclides decay, transmuting into daughter elements that may have different chemical properties. These new elements might degrade the existing mineral by making it unstable. A good waste form that works well for uranium may work poorly for lead, its final decay product. The second problem is that the radioactive decay itself damages the solid over time. Radioactive decay is an energetic process in which ejected particles and the recoiling nucleus disrupt the surrounding atoms. A single alpha-decay event can displace thousands of atoms in the surrounding volume. We know from laboratory measurements that radionuclides are more easily released from radiation-damaged structures than from materials that do not sustain radiation damage. The third problem is that radioactive waste, particularly the high level waste from reprocessing of spent nuclear fuel to extract plutonium and uranium, contains a variety of elements with widely varying chemistry. The waste form must incorporate the radionuclides, as well as non-radioactive elements such as silicon and sodium that are present in the waste stream as a result of waste processing. A number of ceramic waste forms have been developed that minimize these problems and provide a potentially useful host for radionuclides. For ceramics, the mineralogy can be tailored to the waste stream by selecting solid mineral phases with structural sites that can accommodate the waste elements, as well as newly formed radioactive decay elements. Radiation damage can be minimized by selecting mineral phases that allow atoms to renew or regain their original crystalline structure, a process known as annealing. For example, actinide phosphate minerals anneal more readily than actinide silicate minerals. Despite the superior thermodynamic stability of crystalline materials, borosilicate glasses have become the preferred waste forms. One reason is that the processing technologies associated with this glass are believed to be easier to adapt to handling highly radioactive material

Bulk amorphous materials

This is the final report for a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this work was to develop the competency for the synthesis of novel bulk amorphous alloys. The authors researched their synthesis methods and alloy properties, including thermal stability, mechanical, and transport properties. The project also addressed the development of vanadium-spinel alloys for structural applications in hostile environments, the measurement of elastic constants and thermal expansion in single-crystal TiAl from 300 to 750 K, the measurement of elastic constants in gallium nitride, and a study of the shock-induced martensitic transformations in NiTi alloys

Radiation tolerance of ceramics—insights from atomistic simulation of damage accumulation in pyrochlores

We have used molecular dynamics simulations to investigate the effects of radiation damage accumulation in two pyrochlore-structured ceramics, namely Gd2Ti2O7 and Gd2Zr2O7. It is well known from experiment that the titanate is susceptible to radiation-induced amorphization, while the zirconate does not go amorphous under prolonged irradiation. Our simulations show that cation Frenkel pair accumulation eventually leads to amorphization of Gd2Ti2O7, and both anion disorder and cation disorder occur during damage accumulation. Amorphization in Gd2Ti2O7 is accompaniedby a density decrease of about 12.7% and a decrease of about 50% in the elastic modulus. In Gd2Zr2O7, amorphization does not occur, because the residual damage introduced by radiation is not sufficiently energetic to destabilize the crystal structure and drive the material amorphous. Subtle differences in damage accumulation and annealing between the two pyrochlores lead to drastically different radiation response as the damage accumulates

Prediction of irradiation spectrum effects in pyrochlores

The formation energy of cation antisites in pyrochlores (A2B2O7) has been correlated with the susceptibility to amorphize under irradiation, and thus, density functional theory calculations of antisite energetics can provide insights into the radiation tolerance of pyrochlores. Here, we show that the formation energy of antisite pairs in titanate pyrochlores, as opposed to other families of pyrochlores (B = Zr, Hf, or Sn), exhibits a strong dependence on the separation distance between the antisites. Classical molecular dynamics simulations of collision cascades in Er2Ti2O7 show that the average separation of antisite pairs is a function of the primary knock-on atom energy that creates the collision cascades. Together, these results suggest that the radiation tolerance of titanate pyrochlores may be sensitive to the irradiation conditions and might be controllable via the appropriate selection of ion beam parameters

Structure and radiation response of anion excess bixbyite Gd2Ce2O7

International audienc

Function Modelling using the System State Flow Diagram

yesThis paper introduces a rigorous framework for function modelling of complex multi-disciplinary systems based on the System State Flow Diagram (SSFD). The work addresses the need for a consistent methodology to support solution neutral function based system decomposition analysis, facilitating the design, modelling and analysis of complex systems architectures. A rigorous basis for the SSFD is established by defining conventions for states and function definition and representation scheme, underpinned by a critical review of existing literature. A set of heuristics are introduced to support the function decomposition analysis and to facilitate the deployment of the methodology with strong practitioner guidelines. The SSFD heuristics extend the existing framework of Otto and Wood (2001) by introducing a conditional fork node heuristic, to facilitate analysis and aggregation of function models across multiple modes of operation of the system. The empirical validation of the SSFD function modelling framework is discussed in relation to its application to two case studies: (i) a benchmark problem (Glue Gun) set for the engineering design community; and (ii) an industrial case study of an electric vehicle powertrain. Based on the evidence from the two case studies presented in the paper, a critical evaluation of the SSFD function modelling methodology is presented based on the function benchmarking framework established by Summers et al (2013), considering the representation, modelling, cognitive and reasoning characteristics. The significance of this paper is that it establishes a rigorous reference framework for the SSFD function representation and a consistent methodology to guide the practitioner with its deployment, facilitating its impact to industrial practice

Predicting the radiation tolerance of oxides

We have used atomistic computer simulations and ion beam irradiations to examine radiation damage accumulation in multicomponent oxides, We have developed contour energy maps via computer simulations to predict the effects of oxide structure and chemical composition on radiation-induced atomic disorder, defect migration, and swelling. Ion irradiation damage experiments have been perfonned on, pyrochlore and fluorite-structured oxide ceramics to test the predictions from computer models