Miniaturized fracture stress tests for thin-walled tubular SiC specimens
Introduction
Carbon/carbide coated particle fuels have been developed for use in high temperature gas-cooled reactors [1], [2], [3], [4], [5], [6], [7], [8]. The uranium dioxide (UO2) or uranium carbide (UC) fuel kernels are coated with a porous pyrolytic carbon layer (buffer layer), and then with tri-isotropic (TRISO) coatings: inner pyrolytic carbon, silicon carbide (SiC), and outer pyrolytic carbon layers [1], [2], [6], [7], [8]. Among the TRISO layers, the SiC layer is the most important component for the structural integrity of fuel particles because it sustains most of the internal pressure caused by fission gas generation [3], [4], [5].
Since the diameter of the SiC coating is only about 0.9 mm and its thickness is usually less than 0.05 mm, no procedure for testing and evaluation of such a small component has been well established [9]. Further, it is known that the fracture strength of a ceramic material is dependent on the size and shape of the specimen [10], [11], [12]. In this study, therefore, testing and evaluation methods to produce the mechanical property data using miniature SiC specimens were developed and the size effects in small specimens were investigated. This paper introduces two testing and evaluation methods: an internal pressurization method and a diametrical loading method. Test results for chemically vapor deposited (CVD) SiC tubular specimens and for tubular alumina specimens are reported here.
Section snippets
Stress distribution under internal pressurization
Schematics of the two testing methods developed for tubular specimens are given in Fig. 1. Theoretical solutions to calculate fracture stresses using the two loading techniques are described here. If a tubular specimen is internally pressurized to a pressure, P, by compression of an elastomeric insert, the axi-symmetrical hoop stress is expressed as a function of the radial distance from the centerline of the tubular specimen, r [10], [12], [13], [14], [15] as:where ri
Experimental
Tubular SiC specimens were obtained from rod-type surrogate fuels in which the pyrolytic carbon and silicon carbide layers were deposited on graphite rods using the CVD processes. In the process, graphite rods with two different nominal diameters, about 0.9 mm and 1.0 mm, and with a length of 6 mm are coated in a fluidized bed using a vertical, high temperature furnace. The SiC layer is deposited on the pyrolytic carbon layer by decomposition of methyltrichlorosilane (CH3SiCl3) in a hydrogen
Fracture stress of SiC and statistical characteristics
The Weibull plots for fracture stress (variable x) from the two versions of SiC specimens are displayed in Fig. 2. The two testing methods were applied to the specimen SiC-A. The fracture stress by the diametrical loading method was about 18% higher than that by the internal pressurization method; the mean values were 282.6 and 238.8 MPa, respectively. The fracture stress depends on the effective volume or area [10], [11], [12]. In the tubular CVD SiC specimens, the inner surface is assumed to
Summary
- (1)
Two testing methods using internal pressurization and diametrical loading have been applied to evaluate the fracture stress of miniature tubular specimens.
- (2)
Tubular SiC specimens with a wall thickness of about 100 μm and inner diameters of about 0.9 mm (SiC-A) and 1 mm (SiC-B) were tested using the two methods. The mean values of fracture stresses 239, 263, and 283 MPa were measured for SiC-A and SiC-B by internal pressurization, and SiC-A by diametrical loading, respectively.
- (3)
Fracture stress results
Acknowledgements
This research was sponsored by the US Department of Energy under Contract DE-AC05-00OR22725 with UT-Battelle, LLC through a Nuclear Energy Research Initiative Grant.
References (20)
- et al.
J. Nucl. Mater.
(1972/1973) - et al.
J. Nucl. Mater.
(2001) - et al.
J. Nucl. Mater.
(2003) - et al.
Nucl. Eng. Des.
(2003) Carbon
(1965)- et al.
J. Nucl. Mater.
(1972) - et al.
Carbon
(1975) - et al.
J. Nucl. Mater.
(1972) - et al.
Nucl. Tech.
(1977) - et al.
JTEVA
(1991)
Cited by (43)
Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs
2024, Journal of Nuclear MaterialsTransient fuel behavior analysis of UN fuel with a two-layered SiC cladding based on multiphysics method
2023, Nuclear Engineering and DesignRecrystallization and grain growth of Zr-Nb-Sn alloy in 400–500 °C and effect on hydride embrittlement
2023, Journal of Nuclear MaterialsSensitivity analysis applied to SiC failure probability in TRISO modeled with BISON
2022, Progress in Nuclear EnergyEffect of ultra-high temperature treatment on microstructures and mechanical properties of TRISO particles
2022, Journal of Nuclear Materials