Herbert Feinroth

The US has started working to develop a ceramic clad called triplex silicon carbide cladding that can replace zirconium alloys, which could enable power higher uprates and could significantly reduce or eliminate the accidents due to faulty designs such as loss-of-coolant accident (LOCA). Several tests are conducted to assess the structural stability of silicon carbide such as researchers irradiated clad sections in a test loop at the Massachusetts Institute of Technology (MIT) research reactor under specific conditions such as inlet of the loop is maintained at 300°C and 1500 psi. MIT's Center for Advanced Nuclear Energy Systems (CANES) conducted fuel rod analysis to evaluate the ability of the fuel cladding to accommodate high burnups without excessive fuel temperatures. FRAPCON modeling showed that the SiC clad uranium oxide fuel is expected to operate at a higher temperature and release more fission gas. MIT study showed that the gaps of the clad closed at about 70MWd/kgU, after which further swelling of the pellets leads to excessive stresses in the cladding.

An experiment was conducted in the MIT Research Reactor (MITR) to irradiate triplex silicon carbide fuel cladding tubes under typical Pressurized Water Reactor conditions. Measurements were made to determine the impact of exposure on strength and swelling. The Sic clad tubes were fabricated by Ceramic Tubular Products (CTP) with dimensions typical of 15 x 15 commercial PWR reactor fuel. The triplex tubes contain 3 layers, an inner monolithic Sic layer to maintain hermeticity, a central SiCiSiC composite layer to provide a graceful failure mode in the event of an accident, and an outer Sic environmental barrier layer. Clad tubes were exposed to 300 OC pressurized water containing boric acid, lithium hydroxide, and hydrogen overpressure, typical of PWRs. Thirty nine (39) specimens of various types were exposed to coolant, some within the neutron flux region and some outside the neutron flux region. Twenty seven (27) were removed for examination and test after 4 months exposure. Following examination, twenty specimens were reinserted for additional exposure, along with 19 new specimens. The 4 month specimens were weighed and measured at MIT, and some were shipped to Oak Ridge National Laboratory (ORNL) where they were mechanically tested for hoop strength using a polyurethane plug test apparatus. Results were compared with the preirradiation strength and dimensions. Some specimens retained their original strength after exposure, others with a less homogeneous monolith, lost strength.