Thermal Aspects for Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

Abstract

SuperCritical Water-cooled Reactors (SCWRs) are a Generation IV nuclear reactor concept. Two main SCWR design concepts are Pressure-Vessel (PV) type and Pressure-Tube (PT) type reactors. SCWRs would use light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). A reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is that a SCW NPP will have a thermal efficiency of 45 to 50%, a remarkable improvement from the current 30–35%. SCWRs have another added benefits such as a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing an SCW CANDU reactor. This concept refers to a 1200-MWel horizontal pressure-tube type reactor with the following operating parameters: a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Materials and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for a fuel is an enriched Uranium Dioxide (UO2). The industry accepted limit for fuel centreline temperature is 1850°C, and previous studies have shown that the fuel centreline temperature of UO2 pellet might exceed this value at certain conditions. Therefore, a thermal conductivity of the fuel must be sufficiently high to transfer large heat flux within a fuel pellet. Also, a sheath material must withstand supercritical pressures and temperatures inside aggressive medium such as supercritical water, so it should be corrosion-resistant, high-temperature and high-yield strength alloy. In general, sheath materials in various SCWR concepts have a temperature design limit up to 850°C. Uranium Carbide and Uranium Dicarbide are excellent fuel choices as they both have higher thermal conductivities compared to conventional nuclear fuels such as uranium oxide, MOX and Thoria. UC and UC2 are high-temperature ceramics. The sheath material being considered is Inconel 600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. To model a generic SCWR fuel channel, a 43-element bundle string was used. In this paper, bulk-fluid, sheath and fuel centreline temperature profiles together with heat transfer coefficient (HTC) profile were calculated along the heated length of a fuel channel. Also, selected thermophysical properties of various nuclear fuels are listed in the present paper.