A Radial Heat Flow Apparatus For Thermal Conductivity Characterisation Of Cylindrical Samples

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. In this frame, the thermal conductivity (TC) of cladding prototype sections are measured both before and after neutron irradiation with the purpose of better understanding the effects of neutron irradiation on the pyrolytic carbon interphase linking the SiC fibres to the SiC matrix.


Published in:
TRANSACTIONS of the American Nuclear Society, 114, 1226-1228
Presented at:
2016 ANS Annual Meeting, Hyatt Regency New Orleans, New Orleans, Louisiana, USA, June 12-16, 2016
Year:
2016
Publisher:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
Keywords:
Note:
Invited Paper
Laboratories:




 Record created 2017-01-12, last modified 2018-01-28

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