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Spark Plasma Sintering of Commercial Zirconium Carbide Powders: Densification Behavior and Mechanical Properties

Commercial zirconium carbide (ZrC) powder is consolidated by Spark Plasma Sintering (SPS). Processing temperatures range from 1650 to 2100 °C. Specimens with various density levels are obtained when performing single-die SPS at different temperatures. Besides the single-die tooling setup, a double-die tooling setup is employed to largely increase the actual applied pressure to achieve higher densification in a shorter processing time. In order to describe the densification mechanism of ZrC powder under SPS conditions, a power-law creep constitutive equation is utilized, whose coefficients are determined by the inverse regression of the obtained experimental data. The densification of the selected ZrC powder is shown to be likely associated with grain boundary sliding and dislocation glide controlled creep. Transverse rupture strength and microhardness of sintered specimens are measured to be up to 380 MPa and 24 GPa, respectively. Mechanical properties are correlated with specimens’ average grain size and relative density to elucidate the co-factor dependencies.

Zirconium Carbide Produced by Spark Plasma Sintering and Hot Pressing: Densification Kinetics, Grain Growth, and Thermal Properties

Spark plasma sintering (SPS) has been employed to consolidate a micron-sized zirconium carbide (ZrC) powder. ZrC pellets with a variety of relative densities are obtained under different processing parameters. The densification kinetics of ZrC powders subjected to conventional hot pressing and SPS are comparatively studied by applying similar heating and loading profiles. Due to the lack of electric current assistance, the conventional hot pressing appears to impose lower strain rate sensitivity and higher activation energy values than those which correspond to the SPS processing. A finite element simulation is used to analyze the temperature evolution within the volume of ZrC specimens subjected to SPS. The control mechanism for grain growth during the final SPS stage is studied via a recently modified model, in which the grain growth rate dependence on porosity is incorporated. The constant pressure specific heat and thermal conductivity of the SPS-processed ZrC are determined to be higher than those reported for the hot-pressed ZrC and the benefits of applying SPS are indicated accordingly.

Effects of loading modes on densification efficiency of spark plasma sintering: sample study of zirconium carbide consolidation

Abstract Theoretical studies on the densification kinetics of the new spark plasma sinter-forging (SPS-forging) consolidation technique and of the regular SPS have been carried out based on the continuum theory of sintering. Both modelling and verifying experimental results indicate that the loading modes play important roles in the densification efficiency of SPS of porous ZrC specimens. Compared to regular SPS, SPS-forging is shown to be able to enhance the densification more significantly during later sintering stages. The derived analytical constitutive equations are utilised to evaluate the high-temperature creep parameters of ZrC under SPS conditions. SPS-forging and regular SPS setups are combined to form a new SPS hybrid loading mode with the purpose of reducing shape irregularity in the SPS-forged specimens. Loading control is imposed to secure the geometry as well as the densification of ZrC specimens during hybrid SPS process.

High Pressure Burst Testing of SiCf-SiCm Composite Nuclear Fuel Cladding

Silicon carbide fiber reinforced silicon carbide matrix (SiCf-SiCm) composite tube is being developed as an accident tolerant fuel cladding material for light water reactors. The mechanical robustness of the material is of critical importance to ensure that the nuclear fuel and fission products are contained during both normal operation and accident conditions, such as the loss of coolant accident (LOCA) encountered in Fukushima Daiichi Plant. We report the development and application of a high pressure burst testing method to evaluate the mechanical robustness of SiCf-SiCm composite cladding. The internal high pressure is generated using a rubber tubing placed within the SiCf-SiCm sample tube. A unique test rig was designed to seal the ends of the rubber tubing while pressurizing it up to 2,000 bar by hydraulic oil. The expanding rubber tubing confined by the SiCf-SiCm sample tube thus exerts a controlled uniform internal pressure to the SiCf-SiCm sample. The full-field strain distribution of the outside surface of the sample was captured by 3D digital image correlation (DIC) method. The acoustic emission (AE) technique was used to detect damage events during the high pressure burst testing.

Modeling Forced Flow Chemical Vapor Infiltration Fabrication of SiC-SiC Composites for Advanced Nuclear Reactors

Silicon carbide fiber/silicon carbide matrix (SiC-SiC) composites exhibit remarkable material properties, including high temperature strength and stability under irradiation. These qualities have made SiC-SiC composites extremely desirable for use in advanced nuclear reactor concepts, where higher operating temperatures and longer lives require performance improvements over conventional metal alloys. However, fabrication efficiency advances need to be achieved. SiC composites are typically produced using chemical vapor infiltration (CVI), where gas phase precursors flow into the fiber preform and react to form a solid SiC matrix. Forced flow CVI utilizes a pressure gradient to more effectively transport reactants into the composite, reducing fabrication time. The fabrication parameters must be well understood to ensure that the resulting composite has a high density and good performance. To help optimize this process, a computer model was developed. This model simulates the transport of the SiC precursors, the deposition of SiC matrix on the fiber surfaces, and the effects of byproducts on the process. Critical process parameters, such as the temperature and reactant concentration, were simulated to identify infiltration conditions which maximize composite density while minimizing the fabrication time.

Multi Scale Characterization of SiC/SiC Composite Materials

SiC fiber-reinforced SiC matrix composites (SiC/SiC) are under consideration as a structural material for a range of nuclear applications. While these materials have been studied for decades, recently new small scale materials testing techniques have emerged which can be used to characterize SiC/SiC materials from a new perspective. In this work cross section nanoindentation was performed on SiC/SiC composites revealing that both the hardness and Young’s modulus was substantially lower in the fiber compared to the matrix despite both being SiC. Using a Scanning Electron Microscopy (SEM) it was observed that the grain growth of the matrix during formation was radially out from the fiber with a changing grain structure as a function of radius from the fiber center. Focused ion beam machining was used to manufacture micro-cantilever samples and evaluate the fracture toughness and fracture strength in the matrix as a function of grain orientation in the matrix.

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.