Nuclear Technology · 1977 · 224 citations · 30 references
EngineeringMaterials EngineeringMaterials ScienceMaterials CharacterizationMechanical Behavior Of MaterialsNuclear MaterialsCeramic Nuclear MaterialsNuclear Fuel MaterialsRadiation Damage In MaterialsRadiation-resistant MaterialsSilicon MaterialsThermal Properties Of MaterialsDefect ClustersFlexural StrengthNeutron IrradiationNeutron Irradiation EffectsSiliconSilicon CarbideSilicon Carbide (Sic) MaterialsThermal Conductivity
Silicon carbide coatings exhibit varied growth structures, resist oxidation when a silica film forms, and, under neutron irradiation, develop point defect clusters or voids that expand volume and can accelerate fission‑product diffusion. The study reviews pyrolytic silicon carbide properties relevant to fuel particle coatings, summarizing mechanical data such as Young’s modulus, flexural and biaxial strengths, and creep behavior. Neutron irradiation markedly lowers thermal conductivity, while Weibull analysis shows that single‑phase beta‑SiC strength remains largely unchanged but silicon‑rich material experiences significant strength loss.
The properties of pyrolytic silicon carbide (SiC) that are important to its use in nuclear fuel particle coatings are reviewed. The structure of material deposited under different conditions varies in growth features and the constituent phases; the principal effects of neutron irradiation on the structure are to create point defect clusters at irradiation temperatures below ∼1000°C (1273 K) and voids above ∼1000°C (1273 K), with a concomitant volume expansion. The thermal conductivity is greatly reduced by neutron irradiation. Data are available for mechanical properties including Young's modulus, flexural strength, biaxial strength, and creep. Some useful results have followed from application of the Weibull model to strength measurements. The strength of single-phase beta-SiC is barely affected by neutron irradiation, but the strength of material containing excess silicon may be seriously reduced after irradiation. SiC has excellent resistance to oxidation as long as a surface film of silica is maintained. Diffusion of fission products is generally very slow, but it may be increased by structural imperfections.
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