Status of geometry effects on structural nuclear composite properties

structural ceramic composites being considered for control rod applications within the VHTR design. While standard sized (i.e. 150-mm long or longer) test specimens can be used for baseline non-irradiated thermal creep studies, very small, compact, tensile specimens will be required for the irradiated creep studies. Traditionally, it is standard practice to use small, representative test samples in place of full-size components for an irradiated study. However, a real problem exists for scale-up of composite materials. Unlike monolithic materials, these composites are engineered from two distinct materials using complicated infiltration techniques to provide full density and maximum mechanical properties. The material properties may be significantly affected when the component geometry or size is changed. It must be demonstrated that the smaller test samples used in an irradiated study will adequately represent larger composite tubes used for control rod applications. To accomplish this, two different test programs are being implemented to establish that small, flat test specimens are representative of the mechanical response for large, cylindrical composite tubes: a size effect study and a geometry effect study

Interfacial push-out measurements of fully-bonded SiC/SiC composites

The direct measurement of interfacial bond strength and frictional resistance to sliding in a fully-bonded SiC/SiC composite is measured. It is shown that a fiber push-out technique can be utilized for small diameter fibers and very thin composite sections. Results are presented for a 22 micron thick section for which 37 out of 44 Nicalon fibers tested were pushed-out within the maximum nanoindentor load of 120 mN. Fiber interfacial yielding, push-out and sliding resistance were measured for each fiber. The distribution of interfacial strengths is treated as being Weibull in form. 14 refs., 5 figs

Enterphase Integrity of Neutron Irradiated SiC Composites

SiC/SiC composites were fabricated from Hi-Nicalon{trademark} fibers with carbon, porous SiC and multilayer SiC interphases. These materials were then irradiated in the High Flux Beam Reactor with fast neutrons at 260 and 900-1060 degrees C to a dose of 1.1X10{sup 25} n/m{sup 2} corresponding to 1.1 displacements per atom (dpa). Results are presented for bend strength of both non-irradiated and irradiated materials. Within the interphases studied the multilayer SiC interphase material showed the least degradation (8-20%) in ultimate bend stress, while porous SiC underwent the greatest degradation ({approximately}35%). The Fiber matrix interphases are studied with TEM for both nonirradiated and irradiated materials. While no irradiation induced microstructural evolution of the interphase was observed, debonding of the interphase from the fiber was observed for all cases. This debonding is attributed to tensile stresses developed at the interface due to densification of the Hi-Nicalon{trademark} fiber. Residual stress analysis of the fiber matrix interface indicates that the irradiation-induced densification of Hi-Nicalon{trademark} and the volumetric expansion of the CVD SiC matrix cause tensile stresses well in excess of those which can be withstood by these, or any other viable SiC composite interphase

Microstructure of V-4Cr-4Ti following low temperature neutron irradiation

The V-4Cr-4Ti alloys displays excellent mechanical properties, including a ductile-to-brittle transition temperature (DBTT) below - 200 C in the unirradiated conditions. Samples were fission neutron- irradiated in HFBR to a 0.4 dpa dose at 100-275 C. Mechanical tests showed significant irradiation hardening which increased with irradiation temperature. Charpy impact testing also showed a dramatic increase in DBTT on the order of 100 to 350 C. The mechanical property changes are correlated with preliminary results from TEM analysis of the defect microstructure resulting from the low-dose neutron irradiations. TEM of the irradiated material showed a nearly constant defect density of {approximately}1.6x10{sup 23}m{sup -3}, with an average defect diameter of slightly greater than 3 nm