Publication Type:Technical Report
Source:Nuclear Fuel Cycle Program, Volume MIT-NFC-TR-098 (2007)
A proposal for using silicon carbide duplex as fuel cladding in light water reactors is investigated. Thecladding consists of a monolithic inner layer surrounded by a tightly wound fiber-matrix composite. Themonolithic layer retains the volatile fission products while the composite adds strength.Empirical models were developed to describe the physical properties of the composite as a function ofoperating temperature and neutron fluence. The FRAPCON steady-state thermo-mechanical fuel rodmodeling code was used to examine the performance of SiC cladding at high fuel burnup and high powerdensity. A comparison of the behavior of the SiC cladding to the conventional zircaloy claddingdemonstrated that the SiC has superior resistance to creep and mechanical degradation due to radiation oroxidation. However, the lower thermal conductivity of the SiC is an issue, which resulted in significantlyincreased peak fuel temperatures. Mixed UO2-PuO2 fuel was also examined in place of traditional UO2pellets, since this may better resemble transmutation fuels of the future. It was found that the use ofplutonium-bearing mixed-oxide fuels further exacerbates the high fuel temperatures.The performance of SiC cladding was also investigated under reactor transient conditions and comparedto the existing regulatory limits for zircaloy cladding. Both the Loss of Flow Accident (LOFA) and theLarge-Break Loss of Cooling Accident (LBLOCA) were studied using the RELAP thermal-hydraulicscode. During LOFA SiC cladding was found to have a lower, but still acceptable, minimum departurefrom nucleate boiling ratio. For the LBLOCA, the peak SiC cladding temperature remains lower than theregulatory limit between initiation of the break and establishment of stable safety injection coolant flow.