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Hafnium carbide is the most refractory binary compound known. ASTM Manual on Zirconium and Hafnium. ASTM International.
The Symposium on Zirconium in Nuclear Applications was held 21–24 August 1973 in Portland, Ore., and was co-sponsored by the American Society for Testing and Materials (ASTM) Committee B-10 on Refractory Metals and Alloys and by the Institute of Metals Division, The Metallurgical Society of the American Institute of Mining, Metallurgical, and Petroleum Engineers (AIME). Program planning was done jointly by representatives of the ASTM Subcommittee B10.02 on Zirconium and Hafnium and by the TMS-IMD Nuclear Metallurgy Committee of AIME. J.H. Schemel, AMAX Specialty Metals Corp., presided as the symposium chairman, and H.S. Rosenbaum, General Electric Co., served as the symposium co-chairman. The session chairman and co-chairman included: A.L. Bement, Jr., Massachusetts Institute of Technology; A. Lowe, Jr., Babcock & Wilcox Company; M.L. Picklesimer, National Bureau of Standards; P.L. Rittenhouse, Union Carbide Corporation; D.E. Thomas, Westinghouse Electric Company; J. Wahler, Combustion Engineering, Inc.; and C.D. Williams, General Electric Company.
STP551 Zirconium in Nuclear Applications directs its attention to some of the serious problems associated with the design, construction, and operation of nuclear reactors. It provides an exchange of current scientific and technical information from the view points of the producers of zirconium and its alloys, the fabricators of components and systems of nuclear reactors, the operating electrical utilities, and the regulatory government agencies.
Contents
- Zirconium for Nuclear Primary Steam Systems
- Development of a Closed-End Burst Test Procedure for Zircaloy Tubing
- Uniform Ultrasonic Inspection of Fuel Sheath Tubing
- Defect Sensitivity in “Lamb Wave” Testing of Thin-Walled Tubing
- Improved Metallography of Zirconium Alloys
- Determination of Solid Solubility Limit of Hydrogen in Alpha Zirconium by Internal Friction Measurements
- Dual Analysis of Longitudinal and Transverse Zirconium Tensile Stress-Strain Data
- Determination of Complete Plane-Stress Yield Loci of Zircaloy Tubing
- Effect of Thermomechanical Processing and Heat Treatment on the Properties of Zr-3Nb-1Sn Strip and Tubing
- Potential for Improvement of Mechanical Properties in Zircaloy Cold-Rolled Strip and Sheet
- Effect of the Annealing Temperature on the Creep Strength of Cold-Worked Zircaloy -4 Cladding
- Thermomechanical Control of Texture and Tensile Properties of Zircaloy-4 Plate
- Creep Strength of Zircaloy Tubing at 400°C as Dependent on Metallurgical Structure and Texture
- Pilger Tooling Design for Texture Control
- Operable Deformation Systems and Mechanical Behavior of Textured Zircaloy Tubing
- Directionality of the Grain Boundary Hydride in Zircaloy-2
- Use of Ion Bombardment to Study Irradiation Damage in Zirconium Alloys
- Mechanisms of Irradiation Creep in Zirconium-Base Alloys
- In—Reactor Creep of Zr—2.5Nb Tubes at 570 K
- In—Reactor Stress Relaxation of Zirconium Alloys
- High Deformation Creep Behavior of 0.6-in.-Diameter Zirconium Alloy Tubes Under Irradiation
- Suppression of Void Formation in Zirconium
- Deformation of Irradiated Zirconium-Niobium Alloys
- Variation of Zircaloy Fracture Toughness in Irradiation
- Strength and Ductility of Neutron Irradiated and Textured Zircaloy-2
- Plastic Instability in Irradiated Zr-Sn and Zr-Nb Alloys
- Assessment of Fracture Studies on Zircaloy-2 Pressure Tubes
- Irradiation Damage Recovery in Some Zirconium Alloys
- Stress Corrosion Cracking of Zircaloys in Iodine Containing Environments
- Microstructure of the Oxide Films Formed on Zirconium-Based Alloys
- Corrosion and Hydriding Performance of Zircaloy Tubing After Extended Exposure in the Shippingport Pressurized Water Reactor
- Characterization of Zircaloy Oxidation Films
- Fracture of Zircaloy-2 in an Environment Containing Iodine
- Study of Zirconium Alloy Corrosion Parameters in the Advanced Test Reactor
- Effect of Surface Treatment on the Irradiation Enhancement of Corrosion of Zircaloy-2 in HBWR
AUTHOR: Schemel JH, Rosenbaum HS
DATE: 1974
PAGES: 523
FORMAT: Hardcopy
ISBN: 978-0-8031-0757-1
DATE: 1974
PAGES: 523
FORMAT: Hardcopy
ISBN: 978-0-8031-0757-1
Names | |
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Other names | |
Identifiers | |
ECHA InfoCard | 100.031.920 |
EC Number |
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RTECS number | |
UN number | 3178 |
Properties | |
ZrC | |
Molar mass | 103.235 g·mol−1 |
Appearance | Gray refractory solid |
Odor | Odorless |
Density | 6.73 g/cm3 (24 °C)[1] |
Melting point | 3,532–3,540 °C (6,390–6,404 °F; 3,805–3,813 K)[1][2] |
Boiling point | 5,100 °C (9,210 °F; 5,370 K)[2] |
Insoluble | |
Solubility | Soluble in concentrated H2SO4, HF,[1]HNO3 |
Structure | |
Cubic, cF8[3] | |
Fm3m, No. 225[3] | |
a = 4.6976(4) Å[3] | |
Octahedral[3] | |
Thermochemistry | |
37.442 J/mol·K[4] | |
Std molar entropy(S | 33.14 J/mol·K[4] |
−207 kJ/mol (extrapolated to stoichiometric composition)[5] −196.65 kJ/mol[4] | |
Hazards | |
Main hazards | Pyrophoric |
GHS pictograms | [6] |
GHS signal word | Danger |
H228, H302, H312, H332[6] | |
P210, P280[6] | |
NFPA 704 | |
Related compounds | |
Zirconium nitride Zirconium oxide | |
Other cations | Titanium carbide Hafnium carbide Vanadium carbide Niobium carbide Tantalum carbide Chromium carbide Molybdenum carbide Tungsten carbide |
Except where otherwise noted, data are given for materials in their standard state (at 25 °C [77 °F], 100 kPa). | |
verify (what is ?) | |
Infobox references |
Zirconium carbide (ZrC) is an extremely hardrefractoryceramic material,[7] commercially used in tool bits for cutting tools. It is usually processed by sintering.
Properties[edit]
Thermal expansion coefficients of ZrC[2] | |
---|---|
T | αV |
100 °C | 0.141 |
200 °C | 0.326 |
400 °C | 0.711 |
800 °C | 1.509 |
1200 °C | 2.344 |
It has the appearance of a gray metallic powder with cubiccrystal structure. It is highly corrosion resistant. This Group IV interstitial transition-metal carbide is also a member of ultra high temperature ceramics or (UHTC). Due to the presence of metallic bonding, ZrC has a thermal conductivity of 20.5 W/m·K and an electrical conductivity (resistivity ~43 μΩ·cm), both of which are similar to that for zirconium metal. The strong covalent Zr-C bond gives this material a very high melting point (~3530 °C), high modulus (~440 GPa) and hardness (25 GPa). ZrC has a lower density (6.73 g/cm3) compared to other carbides like WC (15.8 g/cm3), TaC (14.5 g/cm3) or HfC (12.67 g/cm3). ZrC seems suitable for use in re-entry vehicles, rocket/scramjetengines or supersonic vehicles in which low densities and high temperatures load-bearing capabilities are crucial requirements.[citation needed]
Like most carbides of refractory metals, zirconium carbide is sub-stoichiometric, i.e., it contains carbon vacancies. At carbon contents higher than approximately ZrC0.98 the material contains free carbon.[5] ZrC is stable for a carbon-to-metal ratio ranging from 0.65 to 0.98.
The group IVA metal carbides, TiC, ZrC, and SiC are practically inert toward attack by strong aqueous acids (HCl) and strong aqueous bases (NaOH) even at 100' C, however, ZrC does react with HF.
The mixture of zirconium carbide and tantalum carbide is an important cermet material.[citation needed]
Uses[edit]
Hafnium-free zirconium carbide and niobium carbide can be used as refractory coatings in nuclear reactors. Because of a low neutron absorption cross-section and weak damage sensitivity under irradiation, it finds use as the coating of uranium dioxide and thorium dioxide particles of nuclear fuel. The coating is usually deposited by thermal chemical vapor deposition in a fluidized bed reactor. It also has high emissivity and high current capacity at elevated temperatures rendering it as a promising material for use in thermo-photovoltaic radiators and field emitter tips and arrays.[citation needed]
It is also used as an abrasive, in cladding, in cermets, incandescent filaments and cutting tools.[citation needed]
Production[edit]
Zirconium carbide is made by carbo-thermal reduction of zirconia by graphite. Densified ZrC is made by sintering powder of ZrC at upwards of 2000 °C. Hot pressing of ZrC can bring down the sintering temperature and consequently helps in producing fine grained fully densified ZrC. Spark plasma sintering also has been used to produce fully densified ZrC.[8]
Poor oxidation resistance over 800 °C limits the applications of ZrC. One way to improve the oxidation resistance of ZrC is to make composites. Important composites proposed are ZrC-ZrB2 and ZrC-ZrB2-SiC composite. These composites can work up to 1800 °C.[citation needed]
References[edit]
- ^ abcLide, David R., ed. (2009). CRC Handbook of Chemistry and Physics (90th ed.). Boca Raton, Florida: CRC Press. ISBN978-1-4200-9084-0.
- ^ abcPerry, Dale L. (2011). Handbook of Inorganic Compounds (2nd ed.). CRC Press. p. 472. ISBN978-1-4398-1461-1.
- ^ abcdKempter, C. P.; Fries, R. J. (1960). 'Crystallographic Data. 189. Zirconium Carbide'. Analytical Chemistry. 32 (4): 570. doi:10.1021/ac60160a042.
- ^ abcZirconium carbide in Linstrom, Peter J.; Mallard, William G. (eds.); NIST Chemistry WebBook, NIST Standard Reference Database Number 69, National Institute of Standards and Technology, Gaithersburg (MD), http://webbook.nist.gov (retrieved 2014-06-30)
- ^ abBaker, F. B.; Storms, E. K.; Holley, C. E. (1969). 'Enthalpy of formation of zirconium carbide'. Journal of Chemical & Engineering Data. 14 (2): 244. doi:10.1021/je60041a034.
- ^ abcSigma-Aldrich Co., Zirconium(IV) carbide. Retrieved on 2014-06-30.
- ^Measurement and theory of the hardness of transition- metal carbides , especially tantalum carbide. Schwab, G. M.; Krebs, A. Phys.-Chem. Inst., Univ. Muenchen, Munich, Fed. Rep. Ger. Planseeberichte fuer Pulvermetallurgie (1971), 19(2), 91-110
- ^Wei, Xialu; Back, Christina; Izhvanov, Oleg; Haines, Christopher; Olevsky, Eugene (2016). 'Zirconium Carbide Produced by Spark Plasma Sintering and Hot Pressing: Densification Kinetics, Grain Growth, and Thermal Properties'. Materials. 9 (7): 577. Bibcode:2016Mate....9..577W. doi:10.3390/ma9070577. PMC5456903. PMID28773697.
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