Supercritical water-cooled nuclear reactors (SCWRs) are a Generation IV reactor concept. SCWRs will use a light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). One reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is to increase the thermal efficiency. The thermal efficiency of existing NPPs is between 30% and 35% compared with 45% and 50% of supercritical water (SCW) NPPs. Another benefit of SCWRs is the use of a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing a pressure tube (PT) type SCWR. This concept refers to a 1200-MWel PT-type reactor. Coolant operating parameters are as follows: a pressure of 25 MPa, a channel inlet temperature of 350°C, and an outlet temperature of 625°C. The sheath material and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for the sheath is a zirconium alloy and the fuel is an enriched uranium dioxide (UO2). The sheath-temperature design limit is 850°C, and the industry accepted limit for the fuel centerline temperature is 1850°C. Previous studies have shown that the maximum fuel centerline temperature of a UO2 pellet might exceed this industry accepted limit at SCWR conditions. Therefore, alternative fuels with higher thermal conductivities need to be investigated for SCWR use. Uranium carbide (UC), uranium nitride (UN), and uranium dicarbide (UC2) are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as UO2, mixed oxides (MOX), and thoria (ThO2). Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at high-temperatures. This paper presents the thermalhydraulics calculations of a generic PT-type SCWR fuel channel with a 43-element Inconel-600 bundle with UC and UC2 fuels. The bulk-fluid, sheath and fuel centerline temperature profiles, together with a heat transfer coefficient profile, were calculated for a generic PT-type SCWR fuel-bundle string. Fuel bundles with UC and UC2 fuels with various axial heat flux profiles (AHFPs) are acceptable since they do not exceed the sheath-temperature design limit of 850°C, and the industry accepted limit for the fuel centerline temperature of 1850°C. The most desirable case in terms of the lowest fuel centerline temperature is the UC fuel with the upstream-skewed cosine AHFP. In this case, the fuel centerline temperature does not exceed even the sheath-temperature design limit of 850°C.

SuperCritical Water-cooled nuclear Reactors (SCWRs) utilize a light-water coolant pressurized to 25 MPa with a channel inlet temperature of 350°C and outlet temperature of 625°C. Previous studies have indicated that uranium dioxide (UO2) nuclear fuel may not be suitable for SCWR use, because the maximum fuel centerline temperature might exceed the industry accepted limit of 1850°C. This research paper explores the use of uranium nitride (UN) as an alternative fuel option to UO2 at SuperCritical Water (SCW) conditions. A generic 1200-MWel Pressure-Tube (PT) -type reactor cooled with SCW was used for this thermalhydraulics analysis. The selected fuel option must have a fuel centerline temperature not higher than the industry accepted limit of 1850°C. Furthermore, the sheath (clad) temperature must not exceed the design limit of 850°C. The sheath and bundle geometry were adopted from previous studies. A single fuel channel was modeled using the UN fuel and an Inconel-600 sheath for several Axial Heat Flux Profiles (AHFPs). Uniform, upstream-skewed cosine, cosine and downstream-skewed cosine AHFPs were used. For each AHFP bulk-fluid, sheath and fuel centerline temperatures, and Heat Transfer Coefficient (HTC) profiles were calculated along the heated length of the channel. The calculations show that the UN fuel maintains a centerline temperature well below the industry accepted limit due to its high thermal conductivity at high temperatures. Therefore, the UN nuclear fuel is a viable fuel option for PT-type SCWRs.

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.

SuperCritical Water-cooled Reactors (SCWRs) are one of six next-generation nuclear-reactor design options under consideration worldwide. These nuclear-reactor design options are included in the major international treaties such as: Generation IV International Forum (GIF) and INternational PROject on innovative nuclear reactors and fuels (INPRO). SCWR coolant is light water, which operates at supercritical pressures and temperatures. Typical SCWR coolant operating parameters are 25 MPa and 350–625°C. These SCWR operating conditions significantly increase the thermal efficiency of a SCW Nuclear Power Plant (NPP) (about 45 – 50%) compared to that of existing NPPs (30 – 35%). Also, SCWRs use significantly higher water parameters than existing water-cooled reactors, because of this they can support hydrogen co-generation. Previous thermal-design fuel-channel option studies for SCWRs have shown that the use of uranium dioxide (UO2) fuel at supercritical water conditions might be unacceptable as the fuel centerline temperature is close to or even exceeds the industry accepted limit of 1850°C. Alternative fuels with a higher thermal conductivity have to be considered. Thoria (ThO2) fuel is a suitable alternative to UO2 due to its higher thermal conductivity. Thoria fuel is beneficial because it complies with the Non-Proliferation Treaty and there are plenty of reserves worldwide. Therefore, ThO2 fuel and its suitability with SCWR use are considered in this paper.

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