The goal of achieving higher thermal efficiency in nuclear power systems, whether fission or fusion based, has invariably led to the study and development of refractory metals, ceramics, and their ...composites. Silicon carbide materials, owing to their favorable neutronic and high-temperature properties, have seen extensive study for over half a century in support of this goal. Currently, our community has a relatively deep understanding of the irradiation effects on this system and has developed irradiation-hardened materials that are currently in use for fission reactor fuels and available as structural composites for next generation reactors. Outside of the nuclear arena SiC has also enjoyed significant development with a wide range of ordinary and high-value product now in use including very high temperature commercial aerospace installations such as turbine engines. The paper presents a brief history of the development of SiC, focused on but not limited to irradiation applications that has led to our present understanding of the system for nuclear application.
Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel ...applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.
The Fe–Cr–Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation ...resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe–Cr–Al alloys has not been fully established. In this study, a series of Fe–Cr–Al alloys with 10–18 wt % Cr and 2.9–4.9 wt % Al were neutron irradiated at 382 °C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2〈111〉 and a〈100〉 were detected and quantified. Results indicate precipitation of Cr-rich α′ is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure–property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α′ precipitates at sufficiently high chromium contents after irradiation.
Silicon carbide (SiC) continuous fiber-reinforced, SiC-matrix composites (SiC/SiC composites) are industrially available materials that are promising for applications in nuclear environments. The ...SiC/SiC composites consisting of near-stoichiometric SiC fibers, stoichiometric and fully crystalline SiC matrices, and the pyrocarbon (PyC) or multilayered PyC/SiC interphase between the fiber and the matrix are considered particularly resistant to very high radiation environments. This paper provides a summary compilation of the properties of these composites, specifically those with the chemically vapor-infiltrated (CVI) SiC matrices, including newly obtained results. The properties discussed are both in unirradiated condition and after neutron irradiation to intermediate fluence levels (most data are for <∼10 displacement per atom) at 300–1300°C.
The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and ...significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (∼90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage, providing insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.
► Overview of recent research on radiation effects in SiC is provided. ► Applications include fuel, fission/fusion structures, and waste stabilization. ► Focus on application of modern materials ...modeling methods. ► Contribution to revolutionary safe fuel for current nuclear energy expected.
Silicon carbide has enjoyed both fundamental study and practical application since the early days of nuclear materials science. In the past decade, with the increased interest in increasing efficiency, solving the real issues of waste disposal, and the constant mission to improve safety of nuclear reactors, silicon carbide has become even more attractive. The purpose of this paper is to discuss recent research that not only strives to understand the remarkable radiation stability of this material, but also the practical application of silicon carbide as waste form and for fission and fusion power applications.
Silicon Carbide Oxidation in Steam up to 2 MPa Terrani, Kurt A.; Pint, Bruce A.; Parish, Chad M. ...
Journal of the American Ceramic Society,
August 2014, Letnik:
97, Številka:
8
Journal Article
Recenzirano
Growth and microstructure of a protective or nonprotective SiO2 scale and the subsequent volatilization of scale formed on high‐purity chemical vapor deposited (CVD) SiC and nuclear‐grade SiC/SiC ...composites have been studied during high‐temperature 100% steam exposure. The environmental parameters of interest were temperature from 1200°C to 1700°C, pressure of 0.1 to 2 MPa and flow velocities of 0.23 to 145 cm/s. Scale microstructure was characterized via electron microscopy and X‐ray diffractometry. The Arrhenius dependence of the parabolic oxidation and linear volatilization rate constants were determined. The linear volatilization rate exhibited a strong dependence on steam partial pressure with a weaker dependence on flow velocity. At high steam pressures, the oxide scale developed substantial porosity, which significantly accelerated material recession. The dominant oxide phase for the conditions studied was cristobalite. The oxidation behavior of SiC/SiC composite was strongly dependent on the state of the surface, specifically whether steam could find easy entry into the material via surface‐exposed interface layers. For the case where these as‐machined interfaces were surface coated with matrix CVD SiC, composite recession was found to be essentially that of high‐purity CVD SiC.
Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural and plasma-facing materials will be exposed to unprecedented conditions of irradiation, ...heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma-facing components such as the first wall and the divertor, tungsten (W) has been selected as the leading candidate material due to its superior high-temperature and irradiation properties, as well as for its low retention of implanted tritium. In this paper we provide a review of recent efforts in computational modeling of W both as a plasma-facing material exposed to He deposition as well as a bulk material subjected to fast neutron irradiation. We use a multiscale modeling approach-commonly used as the materials modeling paradigm-to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions.
The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in ...silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.
The SiC layer integrity in the TRISO-coated gas-reactor fuel particle is critical to the performance, allowed burn-up, and hence intrinsic efficiency of high temperature gas cooled reactors. While ...there has been significant developmental work on manufacturing the fuel particles, detailed understanding of the effects of the complex in-service stress state combined with realistic materials property data under irradiation on fuel particle survival is not adequately understood. This particularly frustrates the modeling efforts that seek to improve fuel performance through basic understanding. In this work a compilation of non-irradiated and irradiated properties of SiC are provided and reviewed and analyzed in terms of application to TRISO fuels. In addition to a compilation and review of literature data, new data generated to fill holes in the existing database is included, specifically in the high-temperature irradiation regime. Another critical piece of information, the strength of the SiC/Pyrolytic carbon interface, was measured and is included, along with a formalism for its analysis. Finally, recommended empirical treatments of the data are suggested.
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